Text

Proposed Tech Specs Changes Supporting Accelerated ABWR Schedule,Chapter 16,Amend 33
	ML20058L922
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Site:	05200001
Issue date:	12/08/1993
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Table 1.

Owners' Group Changes Incorporated into ABWR Technical Specifications Chapter 16, Amendment 33 i

Owners' Group Affected Chanaes l

Chance Packaae No.

and Item No.

BWR-01A C1 3.3.1.1 SSLC Sensor Instrumentation-J C1 B 3.3.1.1 SSLC Sensor Instrumentation-Cl, C2 1.10.8

. SDM Test-Refueling C1, C2 B 3.10.8 SDM Test-Refueling BWR-02 C1, C2, C3, C4 1.1 Definition i

C5, C6 1.3 Completion Time.

C8, C8a B-2.1.1 Reactor Core SLs i

C8, C9 B 2.2.4 RCS Pressure SL C11 3.0 LC0 Applicability i

BWR-03 C1 3.4.7

'RHR-Hot Shutdown _

C1, C3 3.4.8 RHR-Cold Shutdown C4 3.9.7 RHR-High Water Level 1

C4, C5 3.9.8 RHR-Low Water Level Cl B 3.4.7 RHR-Hot Shutdown C1, C3 8 3.4.8-RHR-Cold Shutdown C4 8 3.9.7 RHR-High Water Level C4, C5 B 3.9.8 RHR-Low Water Level BWR-04 C1 (in P&R) 3.1.1 SDM C1 (in P&R)

B 3.1.1 SDM C4 3.1.4 Controol Rod Scram Times BWR-04, C3 Rev. 1 3.1.1 SDM and B 3.1.1, SR 3.1.1.1 Frequency C2 Rev. 1, C3 3.1.2 Reactivity Anomalies, and B 3.1.2, SR 3.1.2.1 Frequency C1 (in P&R)

B 3.1.1 -

SDM Required Actions C.1, D.1, j

D.2, D.3, and D.4 BWR-05 C1, C2, C3, C4, CS, C6, C8 1.1 Definitions C14 1.4 Frequency C7, C9 3.0 LCO Applicability C9 8 3.0 LCOs & SRs C12, C15, C11 Rev. 1

.B 3.0 LCOs & SRs BWR-06 C1, C2 1.1 Definitions C4 3.6.2.1 Suppression Pool-Average Temperature C5 3.6.3.1 Primary Containment Hydrogen

-Recombiners C3 Rev. I 1.3 Completion Times-C3 Rev. I 1.4 Frequency-_

C9, Rev. 1 3.3.6.1-PAM Instrumentation bKh5bOaI bh f

Table 1.

Owners' Group Changes Incorporated into ABWR Technical Specifications, Chapter 16. Amendment 33 l

Owners' Groun Affected Chances Chance Packace No.

and item No.

l BWR-II, C2 2.1.2 Reactor Coolant System Pressure SL C3, C4, C5, C6, C7, C8, C9 B2.1.1 Reactor Core SL BWR-13, C2 3.1.1 SDM, Required Action D.4, E.5 i

C3 3.1.3 Control Rod Operability, Completion Time for Reqired Action B.1 and Bases t

l C4 3.1.1 Control Rod Operability,.

Condition D, and Basis C5 3.1.4 Table 3.1.4-1 Note 2 C7 3.1.5 Control Rod Scram Accumulator,

. Condition D C8, C9, C10 B 3.1.1 SDM C10, C12 B 3.1.3 Control Rod Operability-C10, Cll, Cl3, C14, CIS B 3.1.4 Bases Cll, C17, C18 8 3.1.5 Bases CIO, Cll, C19 B.3.1.7 Bases W0G-24 Cl B 3.7.5 Control Room Habitability Area (CRHA) - Air Conditioning (AC)

System C9 4.0 Design Features WOG-22 C6 B 3.4.4 RCS Pressure Isolation Valve (PIV) Leakage WOG-32 C1 1.3 Completion Times WOG-06 Items 1, 3, 5, 6, 7 5.0 Administrative Controls i

BW0G-01 C1, C2 1.1-Definitions C5 1.2 Logical ' Connectors.

C7, C8, C9 1.3 Completion Times C11 3.0 SR Applicability I

BW0G-02, C3 LC0 3.4.6 RCS Specific Activity.

i 1

I.:

r Definitionc 1.1 i

1 1.0 USE AND APPLICATION 1.1 Definitions l

.____..--NOTE-------------------------------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

t Term Definition l

l ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be.taken under designated Conditions within specified Completion

..o A Times.

90*- c?

AVERACE PLS%R EXPOSURE The AVERACE PLANAR EXPOSURE :h:ll bc pplicable to gg,,o G,

specific pl: n:r height :nd i: equal to th: :::

cf the exp;;urc of all the fac1 red; in the C i.

pecified bundle :t_ th: :pecified height divided by the number of fuel red; in the fuel bandic.

AVERAGE PLANAR LINEAR The APLHGR shall be applicable to a specific HEAT GENERATION RATE planar height and is equal to the sum of the (APLHGR)

LHGRs heat generation rate per unit length of fuel rod for all the fuel rods in the specified bundle l

at the specified height divided by the number of fuel rods in the fuel bundle at the height.

i CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel such that it responds within the necessary range and accuracy to i

specified values of the parameter that.the channel monitors. The CHANNEL CALIBRATION shall encompass i

the entire channel, including the required sensor, l

alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temgerature pd Og detector (RTD)ijUifitifiWYisnisiisstor thermocoup

g of an inplace fhenisij Sahiff34 and n'oFiiiil'Eillnifi3E"6f?i@e reidliing-09 n

Std &

idf6IGN1e devices;in the channel. L'h n: ve r :

15 minutes, making up i

at least 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING The ECCS RESPONSE TIME shall be that time interval i

SYSTEM (ECCS) RESPONSE from when the monitored parameter exceeds its ECCS TIME initiation setpoint at the channel sensor until l

the ECCS equipment is capable of performing its l

safety function (i.e., the valves travel to their l

required positions, pump discharge pressures reach j

their required values, etc.).

Times shall include diesel generator starting and sequence loading i

delays, where applicable. The response time may be measured by means of any series of sequential, I

overlapping, or total steps so that the entire response time is measured.

l ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that RESPONSE TIME time interval from when the monitored parameter p0K exceeds its isolation TiiitiitT6iiisetpoint at the y (A channel sensor until thi~isbliff6n valves travel I

to their required positions. Times shall include diesel generator starting and sequence loading delays, where applicable.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that.the entire.

response time is measured.

(continued).

ABWR TS 1.1-4 10/21/93 l

l

i Definiticns 1.1 l

1.1 Definitions l

LOGIC CHANNEL actuation signals generated in the SLU out to the l

(continued) 2-out-of-2 voter.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all FiljilFed logic components (i.e., all FitifiTFidfrsTifs}and contacts, trip functions, TEST 0 p #g

/,'8; E611d"itate logic elements, etc.) of a logic path, from as close to the sensor as practicable up to, i

but not including, the actuated device, to verify 0PERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the I

l entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power i

RATIO (MCPR) ratio (CPR) that exists in the core.

The CPR is that power in the assembly that is calculated by application of the appropriate correlation (s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in tha reactor vessel.

dwR,0,4 t

OPERABLE-0PERABILITY A system, subsystem, ~ v sygng compone.n.t, or pA device shall be OPERABLE,og haveiOPJRAB1Q H when

& Lf it is capable of performing 'its speclffid safety

~

function (s) and when all necessary attendant instrumentation, controls, displays, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, yljfsT67i, component, or device to perform its specifiE'd safety function (s) are also capable of performing their related support function (s).

OUTPUT CHANNEL An OUTPUT CHANNEL is defined as a set of interconnected components that process outputs from associated LOGIC CHANNELS to produce an identifiable signal that deenergizes scram (continued) i ABWR TS 1.1-6 10/21/93

i i

Definitions i

1.1 1.1 Definitions STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is _the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

l TURBINE BYPASS SYSTEM

-The TURBINE BYPASS SYSTEM RESPONSE TIME consists RESPONSE TIME of two components:

Thetime'ff@initialmovementofthemain l

f a.

turbine si.op valve or control valve until 80%

7 i

f(l.,0 of the turbine bypass capacity is established; g(p and l

The time 'iF@ initial movement of the main b.

f turbine sE6"p valve or control valve until initial movement of the turbine bypass valve.

The response time may be-measured by means of any i

series of sequential, overlapping, or total steps so that the entire response time is measured.

l l

l ABWR TS 1.1-10 10/21/93

Logical Crnnect:rs 1.2 1.0 USE AND APPLICATION l.2 Logical Connectors PURPOSE The purpose of this section is to explain the meaning of logical connectors.

l Logical connectors are used in Technical Specifications (TS) l to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TS are AN_Q and QB. The physical arrangement of these connectors constitutes logical conventions with specific meanings.

t BACKGROUND Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic ~

is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector-in the first level of nesting (i.e., left justified with the number of the Required Action).

The successive levels of logic are identified by additional digits of the Required Action number and by successive indentions of the logical i

connectors.

l

" hen logical connector: Or: u::d t: :t:te : C ndit!:r, enh.

o go G,o l the fir:t level of logic i: u::d, :nd the 1 gie:1 : nn :tcr

-+

P

[

i: left ju:tified with the Condition statement.

i

(

When logical connectors are used to state a C6i@fflin, M

Completion Time, Surveillance, or Frequency,"b~nl~f~lhe first level of logic is used, and the logical connector is left justified with the statement of the[f{r@{yjj, Completion Time, Surveillance, or Frequency.

l l

EXAMPLES The following examples illustrate the use of logical connectors.

1 (continued)--

ABWR TS 1.2-1 10/21/93

Completien Times j

1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.

GsjffjRC6]iliM6@ig(ing)efaQ6j (LCOs) specify minimum BACKGROUND 0

M p7-requirements for ensur safe operation of the unit. The ACTIONS associated with an LCO state Conditions that

. h typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action (s) and Completion Time (s).

DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action.

It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable' not within limits) that requires entering an -

ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LC0 Applicability.

If situations are discovered that require entry into more l

than one Condition at a time within a single.LC0 (multiple Conditions), the Required Actions for each Condition must be

. performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the situation that required entry into the Condition.

re 61 Once a Condition has been entered, subsequent H W 6 66s, subsystems, components, or variables expressed *~Wifis~

Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition.

(continued)

ABWR TS 1.3-15 10/21/93

Completion Times 1.3 1.3 Completion Times U64f ?}y Lb)

DESCRIPTIN However, when a subseouent dih ilon, subsystem, component, (continued) or variable expressed in the"Chidifion is discovered to be inoperable or not within limits, the Completion Time (s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:

a.

Must exist concurrent with the first inoperability; and b.

Must remain inoperable or not within limits after the first inoperability is resolved.

The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:

a.

The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; or b.

The stated Completion Time as measured from discovery of the subsequent inoperability.

The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each ' lVillbij, Gd 'O A d

subsystem, component, or variable expressed in iffi'~

C. C Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications.

The above Completion Time extension does not apply to a Completion Time with a modified " time zero." This modified

" time zero" may be expressed as a repetitive time (i.e.,

"once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery..." Example 1.3-3 illustrates one use of this type of Completion Time. The 10 day Completion Time specified for Conditions A and B in Example 1.3-3 may not be extended.

(continued)

ABWR TS 1.3-16 10/21/93

Completien Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued) 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump A.1 Restore pump to 7 days l

inoperable.

OPERABLE status.

I l

B.

Required B.1 Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action and associated.

ANJ Completion Time not B.2 Be in MODE 4. -

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months met.

1 When a pump is declared inoperabla, Condition A is entered.

i If the pump is not restored to OPERABLE status within MD(p'p}

7 days, Condition B is Disifentered and the Completion Time clocks for Required ActT6W~"B.1 and B.2 start.

If the s

61 inoperable pump is restored to OPERABLE status after

~

if e te equ r d i~~

e

'~

f5^6iHEsGa.

When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump.

LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one. inoperable pump.

The Completion Time clock for Condition A does not stop after LC0 3.0.3 is entered, but continues-to be tracked from the time Condition'A was initially entered.

While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited'and operation continut.d in accordance with Condition A.

(continued)

P ABWR TS 1.3-/t%

10/21/93

Completion Times 1.3 i

1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued) y When one Function X sE65fifi(~fjand one Functicn Y sib'ififiiii i

are inoperable, Condition A and Condition B are concur?Esilly

/

applicable. The Completion Times for Condition A and Condiuta B are tracked separately for each subsfith, starting from the time each sGSsy*'sEfired.

stis was deEl &eT inoperable and the Condition ~EE A separate Q

Completion Time is established for Condition C and tracked from the time the second EstiiI[dislribed in Condition C wasstsi was de D

h[D (p (i.e., the time the situalT6h discovered).

If Required Action C.2 is completed within the specified i

Completion Time, Conditions B and C are exited.

If the b

Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A.

The remaining Completion Time in Condition A is measured from the time the affected isbiysfidi was declared inoperable (i.e., initial entry ihf6~C3 Edition A).

The Completion Times of Conditions A and B are modified by a logical connector, with a separate 10 day Completion Time measured from the time it was discovered the LCO was not met.

In this example, without the separate Completion Time, it would be possible to alternate' between' Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO.

The separate Completion Time modified by the phrase "from

[

discovery of failure to meet the LC0" is designed to prevent indefinite continued operation while not meeting the LCO.

This Completion Time allows for an exception to the normal

" time zero" for beginning the Co,?,nletion Time " clock".

In this instance, the Completion Time " time zero" is specified as commencing at the time the LC0 was initially not met, instead of at the time the associated Condition was entered.

l (continued) l l

ABWR TS 1.3-/ M 10/21/93 l

.)

[

J i

-l ComfetienTimes-

.l 1.3' j

i 1.3 Completion Times

.i EXAMPLES EXAMPLE 1.3-4 3

(continued) l ACTIONS 1

CONDITION REQUIRED ACTION-

' COMPLETION TIME-A. One or more A.1 Restore valve'(s) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months valves to OPERABLE inoperable.

status.

B. Required B.1 Be.in MODE 3..

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months j

Action and 4

associated AND l

Completion-Time not B.2 Be in MODE 4.

36-hours i

l met.

I i

A single Completion Time is used for any number of valves.

l inoperable at the same time. The Completion Time associated I

with Condition A is based on the initial entry into' Condition A and is not tracked on a per valve basis.

t Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking _of separate.

i Completion Times.

Once one of the valves has been restored to 0PERABLE status, the Condition A Completion Time is not reset, but continues from the time t;ie first valve was declared inoperable.. The.

Completion Time may be extended'if the valve' restored to-OPERABLE status was the first~ inoperable valve. The-Condition A Completion' Time may be extended.for.up to i

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provided this does not result.in any subsequent valve' being inoperable for > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

m expires while one or more valves ar[e~EtT1T"' inoperab L's Condition B is entered, ges,1

-(continued) _

9 7-

'l.3/

10/21/93 l

ABWR TS

3 p

l

Completion Times
-i 1.3 i
1.3 Completion Times j
EXAMPLES EXAMPLE 1.3-5 (continued)
ACTIONS j

NOTE-----------------------------

Separate Condition entry is allowed for each -inoperable.
.l valve.
~
j CONDITION
' REQUIRED ACTION COMPLETION TIME' l
i A. One or more-A.1 Restore valve' to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months valves OPERABLE status.
inoperable.
.l
.B.
Required 8.1 Be -in MODE 3.
12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s-
-l Action'and-i associated' NH1 l
Completion Time not B.2 Be in MODE 4.
-36' hours-met.
'i 1
The Note above the ACTIONS. table:is-'a method of modifying Dg how the Completion Time is tracked.
If this-method of; modifyingj@ow the Completion Time is Lt EHMERIMEIHlif(- Q' ~g(~~g~{@4 y
gf only to ny:-[:[{:ji]c CgnditioWth N
appear the C:ndi..
. _.Q j g#sD tht 6g i-1 The Note allows Condition' A to be entered separately for...
H each inoperable valve, and Completion Times. tracked on-a per -
valve basis. When a valve is' declared inoperable,.
Condition A is entered and its Completion Time' starts.- If; subsequent valves are' declared inoperable, Condition A is entered for each valve and separate. Completion Times start ;
I and are tracked for each valve.
'i l(continued)-
1.3'-[M 10/21/93 ABWR TS i
l
,I
I Completion Times 1.3
-l 1.3 Completion Times 1
i EXAMPLES EXAMPLE 1.3-6 (continued)
Entry into Condition A offers a choice between Required Action A.1 or A.2.
Required Action A.1 has a "once per" Completion Time, which qualifies.for the 25% extension, per SR 3.0.2 to each performance after the initial performance.
i s#y%
ThiWalt f,illiBVK6&Fif6fiWiF6 fps 4uTFsdfAEff5EWfils esthsdfindlilthsiinitial f ~rformance?6f!RshijijFidyActl6ff#fi W
Ck isst?6eib41stsd!nithin :thEM1@j8s6sustinisWa17 if-P;s4EIVWAEtToE"A!f"Wf 1 Towed and J6[ijthe extension
%g-o b
~
RiijliTFEd~%Etion: is not met within the Completion Time (plj
-s allowed by SR 3.0.2), Condition B is entered.
If Required Ejs;b I
Action A.2 is followed and the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is not met, Condition B is entered.
If after entry into Condition B, Required Action"A.1 or A.2 ~
[
is met, Condition B is exited and operation may then continue in Condition A.
I I
(continued)
I ABWR TS 1.3-11 10/21/93 q
i
Completien Times l.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued)
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One A.1 Verify affected I hour i
subsystem subsystem l
inoperable.
isolated.
M j
Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months i
thereafter P
AND A.2 Restore subsystem 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to OPERABLE i
status.
i B.
Required B.1 Be_in MODE 3.
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action and associated M
i Completion Time not B.2 Be in MODE 4.
36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months met.
Required Action A.1 has two Completion Times..The I hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter" interval begins upon performance of Required Action A.I.
Ic i
If after Condition A is entered, Required Action A.I.is not p,0 met within either the initial I hour or any subseguent p&9 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months interval from the previous performance (plujJthe extent ion allowed by SR 3.0.2), Condition B is 66tered. The
$6. k o
Completion Time clock for Condition A does not stop after Condition B is entered; but continues from the time t
Condition A was initially entered.
If Required Action A.1 (continued)'
ABWR TS 1.3-26 10/21/93 4
1
Completien Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1,3-7 (continued) is met after Condition B is entered, Condition B is exited g 'o\\
and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not t
expired. Sine th: ::::nd C :pletion Time ef ".: quired bgo 9
Acti:n A.! h:: : : dified "ti : : r;" (i.e., efter the (f
initial I h: r, not frc: ti : Of C nditi:n ::try), th
lle=ne for : C :pleti n Tiz exten ica d::: n t :p;.ly.
IMMEDIATE When "Immediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.
7i ABWR TS 1.3-10/21/93'
Frequency 1.4 1.4 Frequency DESCRIPTION criteria.
SR 3.0.4 restrictions would not apply if both the (continued) following conditions are satisfied:
a.
The Surveillance is not required to be performed; and b.
The Surveillance is not required to be met or, even if required to be met, is not known to be failed.
EXAMPLES The following examples illustrate the various ways that Frequencies are specified.
In these examples, the Applicability of the LC0 (LC0 not shown) is MODES 1, 2, and 3.
EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHANNEL CHECK.
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ) during which the associated Surveillance must be performed at least one time.
Performance of the Surveillance initiates the subsequent g '# (o extension of the time interval to 1.25 times the [(fitEWa,t interval. Although the Frequency is stated as 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months an Q[d }
spssiffsd![1iiWha Frequency is allowed by SR 3.0.2 f6F~~~
6piFifl66Al~fisiibility. The measurement of this interval J
continues at all times, even when the SR is not required to b
be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the l
unit is outside the Applicability of the LCO).
If the interval specified by SR 3.0.2 is exceeded while the unit is in a MODE or other specified condition in the Applicability i
of the LCO, and the performance of the Surveillance is not l
otherwise modified i
(continued) l ABWR TS 1.4-29 10/21/93 1
FrGqusney 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-1 (continued)
(refer to Examples 1.4-3 and 1.4-4), then SR 3.0.3 becomes applicable.
If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a MODE or other specified condition in the Applicability of the LC0 for which performance of the SR is required, the Surveillance must be performed within the-Frequency requirements of.SR 3.0.2 prior to entry into the MODE or other specified condition.
Failure to do so would result in a violation of SR 3.0.4.
EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits.
Once within-12 hours after 2'25% RTP M
24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, ar.d the second is'of the type shown in Example 1.4-1.
The logical connector "M" indicates that both Frequency requirements must be met.. Each time reactor power is increased from a power level < 25% RTP to 2 25% RTP, the Surveillance must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
The use of "once" indicates a' single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by " M ").
This type of Frequency does not qualify for the extension-allowed by SR 3.0.2.
4#/r-6WR'Oh g,S 1 (continued)'
ABWR TS 1.4-30 10/21/93
FrGquency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1,4-2 (continued)
"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example).
If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.
EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

NOTE------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 2 25% RTP.
Perform channel adjustment.
7 days The interval continues whether or not the unit operation is
< 25% RTP between performances.
As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 1
power reaches 2 25% RTP to perform the Surveillance.
The Surveillance is still considered to be within the "specified bg 'ob Frequency." Therefore, if the Surveillance were not 6
performed within the 7 day interval !)]iisJt}IsITifEii5M allowedWi!$R3MQ but operation was < 150RTP it would d
6js. Y h6tl6hitillife"i'fiilure of the SR or failure to meet the 88 LCO. Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided i
operation does not exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with power 2 25% RTP.
(continued)
ABWR TS 1.4-W 10/21/93
Frequ:ncy 1.4
)
i 1.4 Frequency i
1 1
EXAMPLES EXAMPLE 1.4-3 (continued)
Og Once the unit reaches 25% RTP,12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months would be allowed for conipleting the Surveillance.
If.the Surveillance were not
@Pcp/
performed within this 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval, there would then be
/
a failure to perform a Surveillance within the specified 4
Frequency; "00E ch:n;;;; then would be re:tricted in hg accordance with SP, 3.0.4, and the provisions of SR 3.0.3
/tt y
would apply.
J t
V i
EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS i
SURVEILLANCE FREQUENCY _

NOTE------------------

Only required to be met in MODE 1.
1 Verify leakage rates are within limits.
24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Example 1.4-4 specifies that the requirements of this Surveillance do not have to be met until the unit is'in MODE 1.
The interval measurement for the Frequency of this Surveillance continues at all times,-as described in j
Example 1.4-1.
However, the Note constitutes _an "otherwise j
0g stated" exception to the Applicability of this Surveillance.
7 Therefore if the Surveillance were not performed within the Y
kQslihs?sHshif667i1T6Gidj but the [ unit was HafiE^h0DE~'17fnere$}f3!Q)] interval, 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months i
(, 4 would*be no failure of pg the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.4 occurs when changing MODES, even with the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency exceeded, provided the MODE change was not made into MODE 1.
Prior to entering MODE 1.(assuming again that the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency were not met), SR 3.0.4 would require satisfying the SR.
l
-1 j
ABWR TS 1.4 10/21/93
l SLs 2.0 I
?.0 SAFETY LIMITS (SLs) i 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 55.2 Kg/cm*g (785 psig) or core flow < 10% rated core. flow:
'l 1
THERMAL POWER shall'be s 25% RTP.
2 2.1.1.2 With the reactor steam dome pressure 2 55.2 Kg/cm g (785 psig) and core flow 210% rated core flow:
MCPR shall be 21.07.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.:
~
($w V O 0
2.1.2 Reactor Coolant System Pressure SL
/
Reactor steam dome pressure shall be =intained s 93.1 Kg/cm g (1325 psig).
I l
2.2 SL Violations l
With any SL violation, the following actions shall be completed:
2.2.1 Within I hour, notify the NRC Operations Center, in accordance with 10 CFR 50.72.
2.2.2 Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:
2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all insertable control rods.
2.2.3 Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , notify the [ General Manager-Nuclear Plant and i
l Vice President-Nuclear Operations] and the-[offsite reviewers l
specified in Specification 5.5.2, "[0ffsite] Review and Audit"].
i
.(continued) 1 ABWR TS 2.0-1 10/21/93 l
I
l Reactor Cere'SLs B 2.1.1 -
B 2.0 SAFETY LIMITS (SLs)
B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref. 1) requires, and SLs ensure,-that specified' acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (A00s).
g The fuel cladding integrity-SL is set such that no
[p Ejd1Tf6fiif] fuel damage is calculated to occur if the limit -
is nofid61ated.
Because fuel damage is not.directly t
l observable, a stepback approach is used to establish an SL,.
such that the MCPR is not less than the limit specified in Specification 2.1.1.2. MCPR greater than the specified limit represents a conservative margin relative to the conditions -
required to maintain fuel cladding-integrity.
l The fuel cladding is one of the physical barriers that separate the radioactive materials from the environs. :The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although-some corrosion or use~ related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously
^
measurable.
Fuel cladding perforations, however, can result from thermal stresses, which occur from reactor operation significantly above design conditions.
While fission product migration from cladding perforation.is just as measurable as that from use'related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross, rather than incremental, cladding deterioration. Therefore, i
th~e fuel cladding SL is defined with a margin to the -
i conditions that would produce onset of transition boiling 1
(i.e., MCPR - 1.00). These conditions represent a significant departure from the condition intended by design for planned operation. The MCPR fuel cladding integrity SL i
ensures that during normal operation and during'A00s, at least 99.9% of the fuel rods in the core do not experience transition boiling.
l (continued)
ABWR TS B 2.0-1 10/21/93
i Reactor Ccre SLs B 2.1.1 BASES BACKGROUND Operation above the boundary of the nucleate boiling regime (continued) could result in excessive cladding temperature because of the onset of transition boiling and the resultant sharp reduction in heat transfer coefficient.
Inside the steam film, high cladding temperatures are reached,. and a cladding water (zirconium water) reaction may take place. This chemical reaction results in oxidation of the fuel cladding to a structurally weaker form. This weaker form may lose its integrity, resulting in an uncontrolled release of.
activity to the reactor coolant.
I l
APPLICABLE The fuel cladding must not sustain damage _ as a result of SAFETY ANALYSES normal operation and A00s. The reactor core SLs are established to preclude violation of the fuel design criterion that an MCPR liiiifQis to ~be established, such that at least 99.9% of the flii1~ rods in the core would not be d) expected to experience the onset of transition boiling.
a b N 1
The Reactor Protection S
$shib^N16stVGiiistitisj")ystem setpoints (LC0 3.3.1.1, "83f4
, in combination with all the Ltds" Wi~diB5hEd~157FiY5nt any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result in i
reaching the MCPR limit.
.~
2.1.1.1 fuel Claddina Inteority (General Electric 6#1)
c. 6 p lI Comoany (GE) Fuel)
&Y GE critical power correlations'are applicable for all 2
critical power calculations at pressures 2 55.2 Kg/cm g (785 psig) or core flows 210% of rated flow. ' For operation at low pressures and low flows, another basis is used,-as i
follows:
Since the pressure drop'in the bypass region is essentially all elevation head, the core pressure drop at loy (power and flows will always be >
.316.Kg/cm 4.5 psi). Analyses Ref. 2) show-that witp a bundle flow of 12.7 m(/h-l (28 x 10 lb/hr), bundle pressure drop is nearly 4
_(continued)-
ABWR TS B 2.'0-2' 10/21/93 l
Reactor Core SLs B 2.1.1 BASES a
APPLICABLE 2.1.1.1 Fuel Claddina Intearity (General Electric SAFETY ANALYSES Company (GE) Fuel)
(continued)
.independen) {of bundle power and has a value of
.246 Kg/cm 3.5 psi). Thus, the bundle flow with a.316Kp/cm driving head will be
> 12.7 m /Es (28 x 10(A.5 psj)lb/hr).
.FullscalgATLAS 4
test data Tsken at pressupes from 1 Kg/cm a (14.7 psia) to 56.2 Kg/cm a (800 psia) indicate that the fuel assembly critical power at this M.
peakingfactors,thiscorresp)ondstoaTHERMALWith the des S
flow is approximately 3.35 MI(t c~l POWER > 50% RTP. Thus, a THERMAL POWER limit of j
25% RTP for reactor pressure < 55.2 Kg/cm'g (785 psig) is conservative.
j l
2.1.1.2 MCPR (GE Fuel)
The fuel cladding integrity SL is set such that no significant fuel damage is calculated to ' occur if the limit is not violated.
Since the parameters that i
result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions that result in the onset of transition boiling have been used to mark the-beginning of the region in which fuel damage could occur Although it
.is recognized that the onset.of transition boiling would not result in damage to BWR_ fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient
~
limit. However, the uncertainties'in monitoring the core operating state and in the procedures used to i
calculate the critical power result in an uncertainty in the value of the critical power..Therefore, the I
fuel cladding integrity SL is defined as the. critical power ratio in the limiting fuel assembly.for which more than~99.9% of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties.
The MCPR SL is determined using a statistical model that combines all the uncertainties in operating (continued)
ABWR TS B 2.0-3 10/21/93
1 Reacter Ccre SLs B 2.1.1 A
BASES APPLICABLE 2.1.1.2 MCPR (GE Fuel)
(continued)
SAFETY ANALYSES parameters and the procedures used to calculate -
critical power. The probability of the occurrence of boiling transitionLis determined using the approved General Electric Critical Power correlations. Details of the fuel cladding integrity SL-calculation are given in Ref. 2.
Ref. 2 also includes a tabulation of the uncertainties used in the determination of the i
MCPR SL and of the nominal values of the parameters used in the MCPR SL statistical analysis.
2.1.1.3' Reactor Vessel Water level-During MODES 1 and 2, the' reactor vessel water level i
is required to be above the top of the active fuel to.
provide core cooling capability. With fuel in the-1 reactor vessel during periods when the reactor'is ' shut down, consideration must be given to water level.
requirements due to the effect of decay heat.
If the water level should drop below the top of the active-
~
irradiated fuel during this period, the ability to remove decay heat is reduced. This reduction in 1
cooling capability could' lead to elevated cladding -
temperatures and clad perf ion in the event that GaJ 6'll the water level becomes
j BnLea 0f W
C 2.2.5 If any SL is violated, restart of the unit shall not i
commence until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before l
l the unit begins its restart to normal operation.
j REFERENCES 1.
10 CFR 50, Appendix A, GDC 10.
i l
2.
NEDE-240ll-P-A, (latest approved revision).
l 3.
10 CFR 50.72.
4.
10 CFR 100.
5.
10 CFR 50.73.
1 ABWR TS B 2.0-6 10/21/93 1
I.
1
J
\\
l l
RCS Pressure SL j
B 2.1.2 l
BASES (continued)
APPLICABLE The RCS safety / relief valves and the Reactor Protection SAFETY ANALYSES System Reactor Vessel Steam Dome Pressure-High Function have settings established to ensure that the RCS pressure SL will not be exceeded.
The RCS pressure SL has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to ASME, Boiler and Pressure Vessel Code,Section III, [later Edition], including Addenda through the-
[later Edition] (Ref. 5), whiph permits a maximum pressure transient of 110%, p6.7 Kg/cm'g (1375 psig), of design pressure 87.9 Kg/cm g (1250 psig). The SL of 93.1 Kg/cm'g (1325psig),asmeasuredbythereactopsteamdomepressure I
indicator, is equivalent to 96.7 Kg/cm g (1375 psig) at the lowest elevation of the RCS. The RCS is designed to ASME Code,Section III, [later Edition] (Ref. 6), for the reactor recirculation piping, which permits a maximum pressyre transient of 110% of design pressures of 87.lg (g/cm g K
(1250 psig) for suction piping and 116 Kg/cm 1650 psig) for discharge piping. The RCS pressure SL is selected to be the lowest transient overpressure allowed by the applicable codes.
SAFETY LIMITS The maximum transient pressure allowable in the RCS pressure vessel under the ASME Code,Section III, is 110% of design pressure. The maximum transient pressure allowable in the j
RCS piping, vplves, and fittings is 110% of design pressures of 87.9 Kg/pm g (1250 psig) for suction piping and 105.5 Kg/cm g (1500 psig) for discharge piping. The most limiting of these two allowances is the 110% of design i
pressure; therefore, the SL. o.n maximup allowable RCS pressuro is established at 963 Kg/cm g (1375 psig).
APPLICABILITY SL 2.1.2 applies in all MODES; howcVer, in "0DE 5, beccuse the recctor vc :cl hecd closure belt: crc act fully 449htered, it i; unlikely the RCS uculd bc ;;rc::uriced.
6WPil C-9 (c 'tinued)
ABWR TS B 2.0-8 10/21/93
RCS Pressure SL i
B 2.1.2 BASES (continued)
SAFETY LIMIT 2.2.1 VIOLATIONS If any SL is violated, the NRC Operations Center must be notified within I hour, in accordance with 10 CFR 50.72 (Ref. 7).
2.2.2 Exceeding the RCS pressure SL may cause immediate RCS failure and create a potential for radioactive releases in excess of 10 CFR 100, " Reactor Site Criteria," limits (Ref. 4). Therefore, it is required to insert all insertable control rods and restore compliance with the SL within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time ensures that the operators take prompt remedial action.
2.2.3 If any SL is violated, the appropriate senior management of the nuclear plant and the utility shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period provides time for plant operators and staff to take the appropriate immediate action and assess the condition of the unit before reporting to the senior management.
2.2.4 If any SL is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the NRC ifR gL accofdajissiVitNMDTCfRWO!73II Rife 8. ni!E3Effi6tfiiTFs pg 'Eg_ ihallifalsbibs7. phosidedi1MihE(inii&F{'manEjeinenFif~Mi~~pji(f p
~
Tiid~the"utllit Vice President-Nuclear g
g.
Operations inditKsI(6ffilis!y/syTeyifi{ipsilfTi'dlQ b
Spec ipEdf l on [515f2MO.f[sjtelRhti eyf and@d t MJJ. f 3
i 2.2.5 If any SL is violated, restart of the unit shall not commente until authorized by the NRC. This requirement ensures the NRC that all necessary reviews, analyses, and actions are completed before the unit begins its restart to normal operation.
(continued)
ABWR TS B 2.0-9 10/21/93
LC0 Applicability-3.0 3.0 L.'E TING CONDITION FOR OPERATION (LCO) APPLICABILITY LC0 3.0.1 L'COs shall be met' during the MODES or other specified -
l conditions in the Applicability, except as~provided in LC0 3.0.2 and LC0 3.0.7.
LC0 3.0.2 Upon discovery of a failure to meet-an LCO,Ethe Required l
Actions of the associated Conditions shall be met, except as provided in [@lipj[$]ig LC0 3.0.6.
L git If the LCO is met or is no longer applicable prior to-l expiration of the specified Completion Time (s),. completion
\\
of-the Required Action (s) is not required, unless otherwise
\\
stated.
\\
.i i
i LC0 3.0.3 When an LC0 is not met and the associated ACTIONS are not l
me,M,j(@@ltidIACT10($, the unit shall be.plac56'TnTh005an a t
[h or other specffiFd condition in which the LCO is not applicable.' Action shall be initiated within I hour to place the unit, as applicable, in:-
l a.
MODE 2 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ; i
I b.
MODE 3 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months ; and c.
MODE 4 within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .
Exceptions to this Specification are; stated in;the individual Specifications.
Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion.
t of the actions required by LC0 3.0.3 is not. required.-
i gwg -e(
LC0 3.0.3 is @ ylapplicable in MODES 1, 2, and 3.
c.7 l
l LC0 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated ACTIONS-to be entered' permit continued operation in the MODE or other specified condition in the l
(continued);
1 l
ABWR TS 3.0-1 10/21/93 l
l L
1
a LC0 Applicability l
3.0 f
f 3.0 LC0 APPLICABILITY LC0 3.0.4 Applicability for an unlimited period of time. This (continued)
Specification shall not prevent changes in MODES or other specified conditions in the Applicability that-are required s
to comply with ACTIONS.
Exceptions to this Specification are stated in the individual Specifications. These exceptions allow entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered-i allow unit operation in the MODE or other specified condition in the Applicability only for a limited period of time.
LC0 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to 6WPof demonstrate its OPERABILITY the OPERABILITY of other equipment, [@Vi~
lM jilM. This is an i
exception to i.t0'filiblis7@l iMh] tem ] returned to service E0'.~2~~fo r he sys i
C) under administrative control to perform the required l
testing.
4 f
LC0 3.0.6 When a supported system LCO is not met solely due to a
[
support system LC0 not being met, the Conditions and Required Actians associated with this supported system are.
not required to be entered. Only the support system LCO l
ACTIONS are required to be entered. This.is an exception to LCO 3.0.2 for the supported system.
In this event, additional evaluations and limitations may be required in i
accordance with Specification 5.8, " Safety Function.
Determination Program (SFDP)."
If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LC0 in which the loss of safety function exists are required to be entered.
When a support system's Required Action directs a supported i
system to be declared inoperable or directs entry into
~i Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LC0 3.0.2.
1 (continued)-
ABWR TS 3.0-2 10/21/93
. =.
SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3 When the Surveillance is performed within the delay period (continued) and the Surveillance is not met, the LC0 must immediately be declared not met, and the applicable Condition (s) must be entered.
SR 3.0.4 Entry into a MODE or other specified condition in the l
Applicability of an LC0 shall not be made unless the LCO's Surve111ances have been met within their specified i
6No(p'#j Frequency. This provision shall not prevent p :::; through i
ee-teshipy?Thf6 MODES or other specified conditions in 5 '!
tenuiM B i s m$f^3ii red.Atticrgjjjj[ g j g [jp g yj g [ Way C05dlNENi i
ininEACIt0!!S2 1
1 I
1 l
ABWR TS 3.0-5 10/21/93 l
LCOs and SRs B 3.0 BASES (continued)
LCO 3.0.5 LC0 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action (s)) to allow the performance of SRs to demonstrate:
og The OPERABILITY of the equipment being returned to a.
service; er
[,9 1
b.
TheOPERABILITYofotherequipment((5 0
5$$d[lE5$5SSN@$5NE5EO5$D5IM i
The administrative controls ensure the time the equipment is i
returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to l
perform the allowed SRs. This Specification does not provide time to perform any other preventive or corrective maintenance.
An example of demonstrating the OPERABILITY of the equipment being returned to service is reopening a containment isolation valve that has been closed to comply with Required Actions, and must be reopened to perform the SRs.
An example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function:from occurring during the. performance of an SR on another channel in the other trip system. A similar example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the appropriate response during the performance of an SR on another channel in the same trip system.
l LC0 3.0.6 LC0 3.0.6 establishes an exception to LC0 3.0.2 for support systems that have an LC0 specified in the Technical-(continued)'
ABWR TS B 3.0-7 10/21/93 l
I
LCOs and SRs B 3.0 i
BASES SURVEILLANCE SR 3.0.1 (continued)
REQUIREMENTS b.
The requirements of the Surveillance (s) are known.to-be not met between required Surveillance performances.
Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable,
-unless otherwise specified. The SRs associated with a Special Operations LC0 are only applicable when the Special Operations LCO is used as an allowable exception to the requirements of a Specification.
Surve111ances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply.<
j Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE i
status.
]
Upon completion of maintenance, appropriate post ' maintenance
)
testing is required to declare equipment 9PERABLE. This j
includes ensuring applicable Surveillances are not failed -
and their most recent performance is in accordance with SR 3.0.2.
Post maintenance testing may not be possible -in the current MODE or other specified conditions'in the 4
Applicability due to the necessary unit parameters not
{
having been established.
In these situations, the equipment-may be considered OPERABLE provided testing has been satisfactorily completed to. the extent possible and the.
equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed. LSome examples of this process'are:
1 a.
Control rod drive maintenance during refueling that requires scram testing at > [~800TpsiQ However, if
/
other appropriate testing is s^ alls ~fEt'orily completed 1
gA and the scram time testing of SR 3.1.3.4~ is satisfied,-
p 'I the control rod can be considered OPERABLE. This allows startup to proceed to reach [8g~pM to
(-4t y perform other necessary testing.
J (continued)
ABWR TS B 3.0-11 ~
10/21/93
LCOs and SRs B 3.0 i
1 l
BASES l
SURVEILLANCE SR 3.0.2 (continued)
REQUIREMENTS requires performance on a "once per..." basis.. The 25%
i extension applies to each performance after the initial performance. The initial performance of the Required i
Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not. allowing the 25%
extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable equipment in an alternative manner.
The provisions of SR 3.0.2 are not intended to be used g oI repeatedly merely as an operational convenience to extend Surveillagce intervals lotNEWiWiththi6iiMoDiMEM~w[f6 0jg ref6 sling /dntIEvali7 or perf53&cMiilition Tiie intervals b"e"y3nd"th53E^T5EEfffed.
l l
SR 3.0.3
^
SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the sgecified Fre uency.,lh,elI~
/
A dela
) 6##f period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6ha{t;5
~ il Ed Gehi.g.q g.fEheYi Mi.sT1hssy apisTfEi e6m"lo h
g. J (
ehEyifwls Oii'iiiiht ~"Tiiiii
~
n performed in accordance with SR 3.0.2, and not at the time
~
that the specified frequency was not met. This delay period provides adequate time to complete Surveillances that have-been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other-remedial measures that might preclude completion of the Surveillance.
The basis for this delay period includes consideration of unit conditions, adequate planning, availability of j
personnel, the time required to ~ perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being l
(continued) l ABWR TS B 3.0-13 10/21/93
SDM 3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D.
(continued)
D.2 Initiate action to I hour restore secondary containment to OPERABLE status.
AND D.3 Initiate action to I hour restore one standby gas treatment (SGT) i subsystem to OPERABLE status.
AND D.4 Initiate action to I hour restore one isolation valve and associated l
instrumentation to I
OPERABLE _jfsd 13 t<> 4 - / y status in each requ secondary containment cp j
penetration flow path not isolated.
E.
SDM not within limits E.1 Suspend CORE Immediately in MODE 5.
ALTERATIONS except for control rod insertion and fuel assembly removal.
AND (continued) 1 1
ABWR TS 3.1-2 10/21/93
.SDM 3.1.I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E.
(continued)
E.2 Initiate action to Immediately fully insert all insertable control rods in core cells i
containing one or more fuel assemblies.
l 5
E.3 Initiate action to I hour restore secondary containment to OPERABLE status.
E E.4 Initiate action to I hour-restore one SGT subsystem to OPERABLE status.
S E.5 Initiate action to I hour restore one isolation valve and associated i
instrumentation to 1
OPERABLE status in eachfsji1Fid OWOU j
secondiriT66tainment cg penetration flow path i
not isolated.
i l
i ABWR TS 3.1-3 10/21/93 i
I i
l
F SDM 3.1.1 l
SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM is:
Prior to each in vessel fuel a.
2 0.38% Ak/k with the highest worth movement during.
l control rod or control rod pair fuel loading analytically determined; or sequence b.
2 0.28% Ak/k with the highest worth AND N
control rod or control rod pair determined by test.
Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after-criticality following fuel movement Eiffdii 6ug-of IW[fiE{$527 c-3 RT?ssyreivessel g,1 rgplacemg]!$
onconge gag I
t h
ABWR TS 3.1-4 10/21/93
Reactivity Anomalies 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY j
SR 3.1.2.1 Verify core reactivity difference between Once within the monitored core k,,, and the predicted 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after core k,,, is within i 1% Ak/k during reaching Operatica in MODE 1.
equilibrium f
conditions
,1 following 6'd y P T startup after fuel movement l
6,y ffii
$M IIiftliihj[2~__l.
NNMs d
ge i
v pressure:Tvesse onbritp61)fM IN1ajj@hij~
SE 1000 MWD /T thereafter BUFlhi)Rf{j binat H0D111 b
3.1[
10/21/93 ABLT! TS
Contral Rod OPERABILITY 3.1.3 l
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I
i A.
(continued)
A.2

NOTE---------

i Not applicable when less than or equal to I
the low power setpoint (LPSP) of the Rod Control and Information System (RCIS).
1 Perform SR 3.1.3.2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and SR 3.1.3.3 for l
each withdrawn j
t OPERABLE control rod.
AND l
A.3 Perform SR 3.1.1.1.
72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.
Two or more withdrawn B.1 Disarm the associated alH6Ei6 6
control rods stock.
CRD.
d
~~~~
MD B.2 Be in MODE 3.
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l
(continued)
I 1
l ABWR TS 3.1 10/21/93
Central Rod OPERABILITY 3.1.3 ACTIONS (continued)
CONDITION REQUIRED ACTION COMPLETION TIME l
C.
One or more control-

NOTE---------

i rods inoperable for
-1.
Inoperable control i
reasons other than
' rods may be bypassed Condition A or B.
in RAPI in accordance j
with SR 3.3.2.1.6, if required, to allow j
insertion of:
i inoperable control rod and continued operation.
2.
Inoperable control rods with failed; l
motor drives can' only -
.be fullyLinserted by individual scram.
s t
i C.1 Fully insert l 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months inoperable control rod AND l
l t
~
C.2 Disarm the associated.
4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CRD.
l 1
l
\\
i l
D.

NOTE---------

D.1 Restore compliance 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months l
l Not applicable when with GWSR.
i THERMAL POWER j
> 10% RTP.
98 l
D.2 Restore control rod 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
-)
@ or more inoperable to OPERABLE status'.~.
0 control rods not in Op g
compliance with Ganged Withdrawal Sequence i
Restrictions (GWSR).
-l and not separated i
by two or more 1
OPERABLE control rods.
-i (continued),
L 3.1[
10/21/93
]
ABWR TS a
-l l
l~
r Centrol Rod Scraa Times 3.1.4 i
Table 3.1.4-1 Control Rod Scram Times

NOTES------------------------------------

t 1.
OPERABLE control rods with scram-times not within the limits of this Table are considered " slow."
i
/D 2.
Ef~~~
Tiiii61 h{(U Fid?fdfliirsi?ER[C6?sTli fdGiffst IfhTNiir"t'iin~ss~T[~]~T)econds Ts'60; Rid [in_sertion[tiL;' Bi 6BiiEF61 ro j^il rod pos ion are inoperable, in accordance with SR 3.1.3.4, and are not considered " slow."
i i
SCRAM TIMES (a)
I (seconds) i REACTOR REACTOR REACTOR STEAM D04E; STEAM i
STEAMDOM()
PRESSURE (
'I R0D POSITION PRESSURE (D)
PRESSUR PERCENT INSERTION-0 Kg/cn,2g 66.8 Kg g'
73.8 Kg/cm g 2
(%)
(0 psig)
(950 psig)
(1050 psig) l 10 (c)
[
]
[
]
l 40 (c)
[
]
[
]
60
[
]
[
]
l (a) Maximum scram time from fully withdrawn position, based on de-energization of scram pilot valve solenoids as time zero.
S'y i
(b)
For fihishiisdiifs reactor steam dome pressures, the scram time criteria
' [pd are Bit ~6i%iiisd"Sy linear interpolation.
i a
(c)
For reactor steam dome pressure s 66.8 Kg/cm g (950 psig), only 60% rod -
insertion position scram time limit applies.
i 2.
l l
r l
ABWR TS 3.1[
10/21/93 i
'l
.l'h r
P l.
Central' Rod Scrat Accumulatsrs 1
3.1.5-ACTIONS (continued)
CONDITION REQUIRED ACTION
COMPLETION TIME' ij C.
Required Action and C.1 '

NOTE--------

A associated Completion Not applicable if'all i
Timeli Iigiil
. inoperable control-
]
gg not
. rod scramL me..
accumulators are
(
associated with fully.
g, N ted control
(,4
[
^i Place the reactor
.ImmediatelyL,
j mode switch in the shutdown position.'
i l
SURVEILLANCE REQUIREMENTS i
SURVEILLANCE-FREQUENCY i
r
~
SR 3.1.5.1 Verify each control rod scram ~ accumulator
7: days pressure is.2130 Kg/ca'g.
j l
l 4
'}
-'M.
ABWR TS 3.1-/
.'10/21/93.
j y
ll 0
l~
['
4 l
..SDM i
B 3.1.1-l
}
B 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.1 SHUTDOWN MARGIN (SDM)
]
1 BASES i
BACKGROUND SDM requirements are specified-to ensure:
f a.
The reactor. can be~ made subcritical from all operating -
conditions and transients and Design Basis Events;.
b...The reactivity transients associated with postulated:
1 accident conditions ~ are controllable within acceptable-1 limits; and
~
c.-
The reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the-i shutdown condition.
These requirements are satisfied by the control: rods',fas i
described in GDC 26 (Ref. 1), which can compensate for the j
reactivity effects of the fuel and water temperature changes -
i experienced during all operating conditions.
i q
-I APPLICABLE The control rod removal error during refueling accident SAFETY ANALYSES analysis (Ref. 2) assumes the core is subcritical.with'the' highest worth control rod; withdrawn.o The analysis of this reactivity insertion event assumes the refueling tinterlocks:
are OPERABLE when the. reactor is in the. refueling modelofJ operation. These interlocks prevent 1the withdrawal! of more--
1 than one control rod,: or control-rod pair, from the. core:
during refueling.
(Special consideration;and requirements o
for multiple control-rod withdrawal; during refueling are covered in Special Operations LCO 3.10.6,. " Multiple Control Rod Withdrawal-Refueling.") The analysis assumes this-condition is.accepuble since the core will'be: shut down-with the highest worth control rod-or' rod' pair. withdrawn, if.
adequate SDM has been demonstrated.
Prevention or. mitigation of reactivity; insertion events is.
necessary to limit energy deposition in;the fuel to preventL significant fuel damage - which could result in' undue release of radioactivity (::c 5 ::: f r.LCO 3.1.7,'"St:r.dby Lig:id fh/ 3L di l
-(continued) l l'
ABWR TS B 3.1-1 10/21/93
B 3.
M N
/
BASES
/
APPLICABLE Control (SLC) System"). Adequate SDM ensures inadvertent SAFETY ANALYSES criticalities will not cause significant fuel damage.
(continued)
SDM satisfies Criterion 2 of the NRC Policy Statement.
l LCO The specified SDM limit accounts for the uncertainty in the demonstration of SDM by testing. Separate SDM limits are provided for testing where the highest worth control rod or rod pair is determined analytically or by measurement. This is due to the reduced uncertainty in the SDM test when the highest worth control rod or rod pair is determined by l
p1}
measurement. When SDM is demonstrated by calculatio associated with a test (17s Q i!~6Bifffnid$0 CdHE15p M.ns not FifUit p@
T6AdihyYsidUsiiEE), additTEnsl siiiFj'iiiiiEif~SFTddE3 t; thE^~
hiEEifiEd"SD"."Timit liMhElddEdito account for uncertainties in the calculation. "TE~enseie"idequate SDM during the l
design process, a design margin is included to account for l
uncertainties in the design calculations (Ref. 3).
APPLICABILITY In MODES 1 and 2, SDM must be provided because I
subcriticality with the highest worth control rod or rod i
pair withdrawn is assumad in the analysis (Ref. 4).
In MODES 3 and 4, SDM is required to ensure the reactor will be held subcritical with margin for a single withdrawn control rod or rod pair.
SDM is required in MODE 5 to prevent an inadvertent criticality during the withdrawal of a single control rod from a core cell containing one or more fuel assemblie:, or of a control rod pair from loaded core cells during scram time testing.
l l
ACTIONS A.1 With SDM not within the limits of the LCO in MODE 1 or 2, l
SDM must be restored within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .
Failure to meet the l
specified SDM may be caused by a control rod that cannot be inserted. The 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Completion time is acceptable, considering that the reactor can still be shut down, assuming no additional failures of control rods to insert,
(
and the low probability of an event occurring during this interval.
(continued)
ABWR TS B 3.1-2 10/21/93
~
i
~'SDN l
'B 3.1.1 1
- BASES
-)
l ACTIONS
- B.d (continued)
If the SDN cannot be restored, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,-- to prevent -the potential. for further reductions in available SDM (e.g.,-additional. stuck control i
rods). The allowed Completion Time:of 12' hours is-reasonable,- based, on. operating-experience,' to reach MODE 3 from full l power conditions in an orderly manner and'without challenging plant-systems.
fa.1 -
With SDM not'within limits.in MODE 3, the operator _ must:
immediately initiate action to fully insert alltinsertable-control rods. Action must continue until all insertable.
1 control rods are fully inserted. -- This action' results in= the:
least reactive-condition for the~ core.
D.I. D.2. D.3 and D.4 With SDM not within limits in MODE '4,2 the ' operator must immediately initiate action to fully Linsert: all insertable-control rods. This-action results in the least reactive condition for the core. Actions must also be initiated within-1.. hour to provide means for control'of~ potential 4
radicactive releases. This includes ensuring Lsecondary >
g g /J containment (LC0 3.6.4.1, " Secondary Containment") is OPERABLE; at least one Standby Gas Treatment l(SGT) [ik.Wi@_ '
1
~
(LC0 3.6.4.3, " Standby Gas Treatment.(SGT) System")--Is OPERABLE; and at least'one secondary containment isolation valve (LC0 3.6.4.2, " Secondary Containment ~ Isolation Valves.
(SCIVs)") and associated instrumentation (LCO 3.3.6.1,.
" Isolation ~ Instrumentation") are.0PERABLE in each associated i
penetration flow path not ' isolated. This may_be performed-
'i as an administrative ~ check, by examining logs or other j
information,. to determineLif the. components are out.of' i
service for maintenance or.other reasons.
It:is not-necessary to perform the SRs_needed~to demonstrate the.
OPERABILITY of the components.. If, however, any required -
component is inoperable, then it must be '
restored to OPERABLE status.
In this case, SRs may need to be performed to restore the component to OPERABLE status.
Actions must continue until all. required components lare'.
OPERABLE.-
(continued)L ABWR TS B 3.1-3 10/21/93-t--n v
w,,
w-r r-+-
--vve-4 y
a e --
r
SOM B 3.1.1 BASES ACTIONS (continued)
E.1. E.2. E.3. E.4. and E.5 i
l With SDM not within limits in MODE 5,~the operator must l
T immediately suspend CORE ALTERATIONS that could reduce SDM, QP6 e.g., -The :u: pen:ica: re n insertion of fuel in the core or the withdrawal of control rods. Suspension of these I
u./0 j activities shall not preclude completion of movement of a c
component to a safe condition.
Inserting control rods or removing fuel,from the core will reduce the total reactivity and are therefore excluded from the suspended actions.
Action must also be immediately initiated to fully insert l
all insertable control rods in core cells containing one or j
more fuel assemblies. Action must continue until all I
insertable control rods in core cells containing one or more fuel assemblies have been. fully. inserted. Control rods in core cells containing no. fuel assemblies do not affect the reactivity of the core and therefore do not have to be i
inserted.
Action must also be initiated within I hour to provide means' for control of potential radioactive releases. This i
includes ensuring secondary containment (LCO 3.6.4.1) is OPERABLE; at least one SGT subsystem (LC0 3.6.4.3) is OPERABLE; and at least one secondary containment isolation valve (LC0 3.6.4.2) and associated instrumentation (LC0 3.3.6.1) are OPERABLE in each associated penetration flow path not isolated. This may be performed as an administrative check, by examining logs or other-l information, to determine if the components are out of service for maintenance or other reasons.
It is not necessary to perform the SRs needed to demonstrate the OPERABILITY of the components.
If, however, any required component is inoperable, then it must be restored to OPERABLE status.
In this case, SRs may need to be performed to restore the component to OPERABLE status. Actions must continue until all required components are OPERABLE.
(continued)'
ABWR TS B 3.1-4 10/21/93
l l
SDM B-3.1.1 BASES SURVEILLANCE SR 3.1.1.1 l
REQUIREMENTS l
i Adequate.SDM must be demonstrated'to ensure'the reactor can-l be made subcritical from any initial operating condition.
l Adequate SDM is demonstrated by testing before or during the-1 i
first startup.after fuel movement, control rod replacement,.
or shuffling within the reactor pressure vessel. Control.'
rod replacement refers to the~ decoupling. and removal.of a..
l control rod from a core. location, and: subsequent replacement with a new control rod or a. control rod from another core
' location. Since core reactivity will vary during..the cycle:
as a function of fuel' depletion and poison burnup,-the-beginning of cycle (B0C) test must also account for changes j
in core. reactivity during the. cycle. Therefore, to obtain.-
i the SDM, the initial measured value must be increased by an-
{
adder, "R", which is the' difference between the calculated 1 value of maximum core reactivity.during the operating. cycle-and the ' calculated BOC core reactivity.
If the value of R-is negative (i.e., BOC"is the most-reactive point in.thei cycle), no correction to the BOC measured value is' red F6 JRef. 4) L. [
cit 60 Tit gyh.qy thuisDMmmW6fse285!
MaNumnlinfa91_
l 1
cE EP_d2N10-
-l The SDM may be, demonstrated during an in sequence" control'.
)
rod pair withdrawal, in which the highest worth control rod :
pair is analytically determined,.or during local' criticals, l
where the highest worth control rod pair is-determined by' testing. Local critical. tests. require the withdrawal: of out of sequence-control Lrods. This testing would'therefore require bypassing of the' Rod Worth Minimizer to allow the l
out of sequence withdrawal, and therefore additionalc.
-requirements must be met-(see LCO 3.10.7, " Control-Rod Testing -Operating").
i The Frequency of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reaching. criticality is allowed to provide a reasonable amount of time to perform
l the required calculations and appropriate' verification.
- During MODE 5,, adequate SDM is 'also' required t'o ensure' the-reactor does not reach criticality during control rod' withdrawals. An evaluation of each-in' vessel fuel movement -
1 i
during fuel: loading (including shuffling fuel within theJ core) :h:ll be perf:;xd $Qejiig@ to ensure adequate.SDM_
$ w a-/3 i
g fp.
(continued)
ABWR TS
.B 3.1 -10/21/93 l
-)
(.
SDM B 3.1.1 i
BASES SURVEILLANCE SR 3.1.1.1 (continued)
REQUIREMENTS is maintained during refueling. This evaluation ensures the intermediate loading patterns are bounded by the safety analyses for the final core loading pattern.
For example, bounding analyses that demonstrate adequate SDM for the most l
reactive configurations during the refueling may be performed to demonstrate acceptability of the entire fuel aarglWiftMthisSDMTHiiRS65@$~~alf5Iij]Q1h"dj 65 g#
movement _l!!guence.
se additfona gsy6ffit6t[6septajffiefsf{41m traccountdo63E t-9 M1 liiid~^oTM13^aT~~
Fe~c^iusnTss"innsiliitly~iadsfy the SR, provided the fuel:
assemblies are reloaded in the'same configuration analyzed j
for the new cycle. Removing fuel from the core will always result in an increase in SDM.
REFERENCES 1.
10 CFR 50, Appendix A, GDC 26.
2.
ABWR SSAR, Section 15.4.1.
3.
ABWR SSAR, Section 4.3.2.
4.
NDE-24011-P-A-9, "GE Standard Application for Reactor-
~
Fuel," Section 3.2.4.1, Sept.1988.
i l
l l
l a
l
'ABWR TS B 3.1-6 10/21/93 1
i 1
l
.c
l i
Reactivity Anomalies B 3.1.2 4
j.
BASES ACTIONS Al (continued) i conditions to determine their~ consistency-with input to design calculations. Measured core and process parameters are also normally evaluated to determine that they are within the bounds of the safety analysis, and safety analysis calculational models may be reviewed to verify that they are adequate for representation of the core conditions.
The required Completion Time of 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s-is based on the low probability of a DBA during this period, and allows sufficient time to assess the physical condition of the reactor and complete the evaluation of the c > 9 design and safety analysis.
i 1
If the core reactivity cannot be restored to within the 1% Ak/k limit, the plant must.be brought to. a MODE'in which the LC0 does not apply. To. achieve this status, the plant must-be brought to at least. MODE 3 within.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is' reasonable, based on i
operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging--
i plant systems.
4 i
SURVEILLANCE SR 3.1.2.1 REQUIREMENTS 4
Verifying the reactivity difference-between the monitored and predicted core k is within the limits of the LCO provides further ass,u,r,ance that plant operation is 1
maintained within the assumptions of the DBA and~ transient j
analyses. The Core Monitoring System calculates the core k
for the reactor. conditions _obtained.from plant ins #trumentation.
A comparison of_the monitored core'k,,,d-to the predicted core k 'ivity difference.
at the:same cycle exposure _is use to calculate the reaUt Thir :::p:ri:! n requirc; the core t b; Oper: ting :t power 107:1 which-4
=ini=iz th: ::: rt:intic: :nd ::::ure;;nt error:, in Order
{te*p 6[D ).
t Obtain :::ningful rc: ult:. Ther:fsrc, th : ;p:ri:ica i:-
3 Only d:n when in.".00E 1.
The comparison'is required when i
the core reactivity has potentially changed by a significant amount. This may occur following a refueling in which new fuel assemblies are loaded, fuel assemblies are shuffled 4
4 (continued)
B3.1-[10 ABWR TS 10/21/93 i
4
Reactivity Anomalies B 3.1.2 BASES SURVEILLANCE SR 3.1.2.1 (continued)
REQUIREMENTS within the core, or control rods are replaced or shuffled.
Control rod replacement refers to the decoupling and removal of a control rod from a core location, and subsequent replacement with a new control rod or a control rod from another core location. Also, core reactivity changes during the cycle. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months interval after reaching equilibrium conditions following a startup is based on the need for equilibrium xenon concentrations in the core, such that an accurate comparison between the monitored and predicted core k # values can be made.
For the purposes of this SR,,'the reactor is assumed to be at equilibrium conditions when steady state operations (no control rod movement or core flow changes) at 2: 75% RTP have been obtained. The 1000 MWD /T Frequency was developed, considering the relatively slow change in core reactivity with exposure and operating, experience related to variations in core reactivity. ThlRcompaWi6hFeijiiffssl{fh'd 1.
lisi^6iisiFifisyfiffpoGsnKsjTel Mwh@hysisimizisths""~f~dfril@
/ M,01 uhitbrtiintifessihdfsiesiOFEsshi!? Errors $nfo$sMi3I6EfisT indhisdfulifssu11snTjeMfoM6dhELcMilidilolitigph}yJ[ M '
pa.4 Me9M!!0_0Ellf
,p V
i REFERENCES 1.
10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.
2.
ABWR SSAR, Chapter 15.
ABWR TS B 3.1 10/21/93
l 1
Centrol Rod OPERABILITY B 3.1.3 BASES ACTIONS A.I. A.2. and A.3 (continued) a reasonable time to test the control rods, considering the potential for a need to reduce power to perform the tests.
Required Action A.2 is modified by a Note that states the requirement is not applicable when below the actual low power setpoint (LPSP) of the RC&IS, since the notch insertions may not be compatible with the requirements of rod pattern control (LC0 3.1.6) and the RCIS (LCO 3.3.1.2).
To allow continued operation with a withdrawn control rod stuck, an evaluation of adequate SDM is also required within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Should a DBA or transient require a shutdown, to preserve the single failure criterion an additional control rod would have to be assumed to have failed to insert when required. Therefore, the original SDM demonstration may not be valid. The SDM must therefore be evaluated (by messurement or analysis) with the stuck control rod at its stuck position and the highest worth OPERABLE control rod psiEissib.Efi_tsd_iWi_l_h?t.h~EEssiEsi.M_C_O assumed to be fully g
With a single control rod stuck in a withdrawn position, the remaining OPERABLE control rods are capable of providing the required scram and shutdown reactivity.
Failure to reach MODE 4 is only likely if an additional control rod adjacent to the stuck control rod also fails to insert during a.
required scram. Even with the postulated additional single Eshff5.1FF6d.if6E.iftnsTsissiflCUJ to in(siEf7sHfffE failure of an adjacent control rod (siidlifs?Ii53Efifid
. M~~
f6 ^ o reach and maintain MODE 3 conditions (Ref. 6).
Required action A.2 performs a indissi6iitest on each remaining withdrawn control rod to eHI6FE~that no additional control rods are stuck.
Therefore, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to perform the analysis to test in Required Action A.3.
f B.1 and B.2 With two or more withdrawn control rods stuck, the stuck ho.ntrol rods should be isolated from scram pressure within $
co urs and the plant brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
f,@4 CRDiaridisusoundi$fdelfssin61 iests.hshlldfiddimioic0M lI' sol atindfiheic6nIF6TIF6diffbisfsEFiETpFsyisfsTdiisihift37Uid h/
The];cinttb1M6d!ca@Qsilitsdjg6%@andbfjj ijo]atjjglig (continued)
ABWR TS B 3.1-5 10/21/93
4 Centrol R:d OPERABILITY B 3.1.3 i
BASES ACTIONS B.1 and B.2 (continued) sii5EfiGDifdMUTEfE6Ef5IljilifiltTSTiT6%5s3Ii&YTMJ#fHCU ih6Fe ~1hiH~Ehi"Ennrol'~~f6d~'s'tEMWilhh:i:irqbsMnditiduallyjisalited rawn position increases the probability that the reactor cannot be shut down if required.
Insertion of all insertable control rods eliminates the possibility of an additional failure of a control rod to insert. The allowed Completion Time' of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
C.1 and C.2 With one or more control rods inoperable for reasons other than being stuck in the withdrawn position, operation may continue, provided the control rods are fully inserted j
within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and disarmed (however, they do not need to be isolated from scram) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .
Inserting a control i
rod ensures the shutdown and scram capabilities are not adversely affected. The control rod is disarmed to prevent inadvertent withdrawal during subsequent operations. The control rods can be disarmed by disconnecting power to the motor drive or by placing the rod in RC&IS IN0P Bypass.
Required Action C.1 is modified by a Note that allows control rodt to be bypassed in the RC&IS if required to j
allow insertion of the inoperable control rods and continued operation. Also, as noted, control rods declared inoperable with a failed motor drive can only be inserted by scram.
Control rods with failed motor drives are not inoperable for this reason alone, but must be considered so upon failure of SR 3.1.3.2 or SR 3.1.3.3, or when not in compliance with GWSR (see LC0 3.1.6).
This does not conflict with SR 3.0.1 since the ability to move the control rod via the FMCRD, as discussed in the bases for SR 3.1.3.2 and SR 3.1.3.3,:is required to prove that the rod is not stuck.
Likewise, loss of position indication, assuming no rod inovement, would not result in control rod (s)-inoperability until failure of SR 3.1.3.1.
SR QQQi6 provides additional requirements when the control.r(ods are bypassed to ensure compliance with the' RWE analysis.
The allowed Completion Times are reasonable, considering the small number of allowed inoperable control rods,. and provide 4
(continued)
ABWR TS B 3.1-6 10/21/93 d
Control Rod OPERABILITY B 3.1.3 i
BASES ACTIONS C.1 and C.2 (continued) time to insert and disarm the control rods in an orderly manner and without challenging plant systems.
1 0.1 and D.2 Out of sequence control rods may increase the potential during a RWE and therefore, the distribution of inoperablereactiv 6 o ganged withdrawal sequence restrict [t%10% RTP, the generic control rods must be controlled. h ions (GWSR) (which is equivalent to previous banked position withdrawal sequence (BPWS) analysis (Ref. 6) requires inserted control rods not l
in compliance with GWSR to be separated by at least two 6) l OPERABLE control rods in all directions, including the Nd diagonal.
Therefore, if enefss or more inoperable control rods are not in compliance w'ilh GWSR and not separated-by at least two OPERABLE control rods, action must be taken to restore compliance with GWSR or restore the control rods to l
OPERABLE status. A Note has been added to the Condition to clarify that the Condition is not applicable when > 10% RTP since the GWSR is not required to be followed'under these conditions, as described in the Bases for LC0 3.1.6.
L1 If any Required Action and associated Completion Time of Condition A, C, D, or E are not met or nine or more inoperable control rods exist, the plant must be brought to l
a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE-3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
This ensures all insertable control rods are inserted and places the reactor in a condition that does not require.the active function (i.e., scram) of the control rods. The number of control rods permitted to be inoperable when l
operating above 10% RTP (i.e., no CRDA considerations) could be more than the value specified, but the occurrence of a large number of inoperable control rods could be indicative of.a generic problem, and investigation and resolution of the potential problem should be undertaken. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on.
operating experience, to reach MODE 3 from full-power (continued) j ABWR TS B 3.1-7 10/21/93
]
+
1 Centrol R:d OPERABILITY z f
B 3.1.3 I
BASES l
ACTIONS E.d (continued)
[
conditions in an orderly manner and without: challenging.
plant systems.
l i
~i SURVEILLANCE SR 3.1.3.1 REQUIREMENTS
(
The. position.of each control rod must' be determined, to-ensure adequate information on control' rod ' position is available to the operator:for: determining. CRD OPERABILITY i
and controlling rod patterns.
Control rod position may be o
determined by the use of. OPERABLE position indicators, by.
moving control rods to. a position ~ with an OPERABLE indicator, or by the use of other appropriate methods. The-1 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this SR'is based on operating i
experience related to expected changes in control rod position and the availability of. control rod position indications in the control room.
i
l SR 3.1.3.2 and SR' 3.1.3.3 j
1 Control rod insertion capability is demonstrated.by' 1
inserting each partially or fully withdrawn control' rod.at j
two notches and observing that.the ' control-rrod moves'.
The-1 control rod may then be returned to its: original position..
p This ensures the control rod is not stuck and is free:to 0
insert on a scram signal.
These Surveillances'are' noti
\\
required when THotMALIPdWEGEQTHIsidiTnTdE@Y h
the actual LPSFIif"thi*RCIS^TiKEe~~tfiE~T6fMiTns not be compatible with the requirements of. rd ;::tterr..
cr.tr:1MisGWSn 7 day F'rlijUsiiEy o(LC0 3.1.6) 'and the RCIS '(LCO 3.3.2.1). ~ The :
^
f SR 3.1;3.2 1s based on operating-1 experience related to the changes =in CRD performance.and the ease of performing notch testing for' fully withdrawn control rods. Partially withdrawn control rods 1are tested at a.
31 day Frequency, based on. the potential: power reduction..
j required to allow the controlJrod

movement,. and considering
~
the large. testing - sample of SR 3.1'.3.2.
Furthermore, the 31 day Frequency:. takes into account.' operating experience
.related to changes in CRD' performance. At'any-time,ifial control rod is immovable, a determination of that controli j
q rod's trippability (OPERABILITY) must be.made and~
appropriate action must be.taken.
(continued)"
ABWR TS B 3.1-8.
10/21/93L
f CCntrol Rod Scram Times B 3.1.4 BASES LC0 indication. The reed switch closes (" pickup") when the (continued) hollow piston passes a specific location and then opens
(" dropout") as the index tube travels upward.
Verification of the specified scram times in Table 3.1.4-1 is accomplished through measurement of the " dropout" times.
To ensure that local scram reactivity rates are maintained within acceptable limits, no more than two of the allowed
" slow" control rods may occupy adjacent locations.
Table 3.1.4-1 is modified by two Notes, which. state control rods with scram times not within the limits of.the Table are considered " slow" and that control rods with scram times
>[
] seconds to 60% rod insertion position are considered inoperable as required by SR 3.1.3.4.
h A
ThiUCCd?ipiillEQihlyftF6FER4slEE6ntF61EFBdifffnEE p
finoliepablei:cuntpolphodnsillyslinsbFtsd?Asdidishnnid y @h l(LC0i31li3)iOSloMssrahminscontidiffodMinaysbe~
conseFyativhi 5.510A$Idd $ jfdeclapediisbpaFableiandinottacEduhfddif5E ii
~
~~
~
~~
APPLICABILITY In MODES 1 and 2, a scram is assumed to function during transients and accidents analyzed for these plant conditions. These events are arsumed to occur during startup and power operation; therefore, the scram function of the control rods is required during those MODES.
In
\\g MODES 3 and 4, the control rods are caly allcwed te 50 wMhdrcwn under Specic1 Operatien ; LC0 3.10.3, " Single t
[t c--._-,
n-a o u t a _,..,,
o,.
e u... a _.-.
.a ien e in,
D I5S!585Ntrbi33"Eith[r$w55"355U'Sh52cb5"~$hi$$$6E S
t U
sblh$i3Ib5TsiihdFiWKlfsiHEsitthiTFFRE5ifis6dsTs?TEEhitsB r Sh0tdowsandfaic6htrolnodibl6Mistapplied EdEijdifE'FEijeffEishtTf6F"E6?itF61"Fid s"c'F~ini~QThisppoifdii capihil fii~~~~
during these conditions.
Scram requirements in MODE 5 are contained in LC0 3.9.5, " Control Rod OPERABILITY -Re fueling. "
(continued)
ABWR TS B 3.1-3 10/21/93
Ccntrol Rod Scram Times B 3.1.4 BASES ACTIONS A.,.1
'D thi?fFifiTil When the requirements of this LC0 are not met,iifhotTS(~
EspifiVs?FiaEtRitillinssFlish?d5EThij;iisE?iiiitin 8
ki thi risths[asssspti 6nsN fsthnisifetyjlshs1fssinsThsfsfoFi
(/
thTiilliif mssrbsliF5irdhTT6~TMODE 6%hiEh~1hTR0'H6is not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE The four SRs of this LC0 are modified by a Note stating that REQUIREMENTS during a single or pair control rod scram time surveillance, the CRD pumps shall be isolated from the associated scram accumulator. With the CRD pump isolated (i.e., charging valve closed), the influence of the CRD pump head does not affect the single or pair control rod scram times. During a full core scram, the CRD pump head would be seen by all control rods and would have a negligible effect.on the scram insertion times.
SR 3.1.4.1 The scram reactivity used in DBA and transient analyses is based on assumed control rod scram time. Measurement of he scramtimeswithreactorsteamdomepressure266.8Kg/cm}g (950 psig) demonstrates acceptable scram times for the transients analyzed in References 2 and 3.
Scram insertion times increase with increasing reactor pressure because of the competing effects of reactor steam dome pressure and stored accumulator energy. -Therefore, demonstration of adequate scram }g (es at reactor steam dome im pressure greater than 66.8 Kg/cm 950 psig) ensures that the scram times will be within the specified limits at higher pressures.
Limits are specified as a function of reactor pressure to account for the sensitivity of the scram insertion times with pressure and to allow a range of pressures over which scram time testing can be performed.
To ensure scram time testing is performed within a reasonable time following a refueling or after a shutdown
t 120 days, all control rods are required to be tested (continued)
ABWR TS B 3.1-4 10/21/93
Control R:d Scraa Times B 3.1.4 BASES n
p6 J SURVEILL".NCE SR 3.1.4.1 (continued) b g\\o REQUIREMENTS (continued) before exceeding 40% RTP fellew:ng
hutdown. This Frequency is acceptable, considering the additional surveillances performed for control rod OPERABILITY, the frequent verification of adequate accumulator pressure, and the required testing of control rods affected by work on control rods or the CRD System.
SR 3.1.4.2 Additional testing of a sample of control rods is required to verify the continued performance of the scram function during the cycle. A representative sample contains at least 10% of the control rods, with. ThETiih M 255fthi FipfsiisfifiVEIilf_no more than 2006f"the cohfF61" rods in thi'siiiible*tistighfifdstidiiljidity26i " slow. "
If more than 9
20% of the sample li~didlared f6"be "'ilow" per the criteria
\\
in Table 3.1.4-1, additional control rods are tested until k
this 20% criterion [ejjW2K6fsihlihilFsTilli@Ti?iHsf is
\\
satisfied er Re uired,tifiET.Ti;;udCS~c^ HEE.6F%tfil Db f6fi1IH55$iFTBfds15WE6EffBFF58i{tEF5iisH55fithikaFs)jih3
~~
fromisl lisahV_eil.@l ssheigskesid n thss LC0Qi ni tti"76F~iil anned g g-gg fid~Y6F^ihi'^Isiiiple should be different for each test. Data from inadvertent scrams should be used whenever possible to avoid unnecessary testing at power, even if the control rods with data were previously tested in a sample. The 120 day Frequency-is based on operating experience that has shown control rod scram times do not significantly change over an operating cycle. This Frequency is also reasonable, based on the additional Surveillances done on the CRDs at more frequent intervals in accordance with LC0 3.1.3 and LCO 3.1.5,
" Control Rod Scram Accumulators."
SR 3.1.4.3 When work that could affect the scram insertion time is performed on a control rod or the CRD System, testing must be done to demonstrate that each affected control rod retains adequate scram performance over the range of applicable reactor pressures from zero to the maximum permissible pressure. The scram testing must be performed once before declaring the control rod OPERABLE. The (continued)
ABWR TS B 3.1-5 10/21/93
Centrol Rnd Scram Accumulators B 2.1.5 BASES APPLICAB8E Control rod scram accumulators satisfy Criterion 3 of the SAFETY Ai 'YSES NRC Policy Statement.
(continued) l l
[
LC0 The OPERABILITY of.the control rod scram accumulators is required to ensure that adequate scram insertion capability-exists when needed over the entire range of reactor pressures. The OPERABILITY of the scram accumulators is based on maintaining adequate accumulator pressure.
APPLICABILITY In MODES I and 2, the scram function is required for l
mitigation of DBAs and transients and, therefore, the scram accumulators must be OPERABLE to support the scram function.
I In MODES 3 and 4, control rods are only allowed to be h,h withdrawn under S:cci:1 Oper:tions LCO 3.10.3, "C:ntrol R0d Uithdr:wal " t S;utdown," :nd LC0 3.10.t "C ntr:1 R:d l
[g0 "ithdrawal--Cold Shutdown " which rovidcIIKEsithijiWeigj@
g\\
(56dii%itEQQ$iSKEfg[#1[$d[i]{p[ntFXoappllpysflh t
fjd&1psg c
j ontrol l
rod scram accumulif6r OPERABILITY under these conditions.
Requirements for scram accumulators in MODE 5 are contained in LC0 3.9.5, " Control Rod OPERABILITY-Refueling."
ACTIONS The ACTIONS table is modified by a Note indicating that a separate Condition entry is allowed for each control rod scram accumulator. This is acceptable since the Required i
Ag Actions for each Condition provide appropriate compensatory action for each inoper:bleifffffsd control. rod. Complying D[&d withtheRequiredActionsiiij'[ilT6wforcontinuedoperation andsubsequentincperabicI[fic{e3controlrodsgovernedby
(
subsequent Condition entry and application'of associated i
Required Actions.
AJ I
With one control rod scram accumulator' inoperable, the scram function could become severely degraded because the accumulator is the primary source of scram force for the associated control rod or rod pair at all reactor pressures.
(continued) i ABWR TS B 3.1-2 10/21/93
Centrol Rod Scraa Accumulators B 3.1.5 i
BASES l
y ACTIONS L1 (continued)
I modified by a Note stating that.the. Required-Action is 'not-applicable if all control rods associated with the inoperable scram accumulators are _ fully inserted, since the function of the control rods has been performed.
SURVEILLANCE SR 3.1.5.1-t REQUIREMENTS SR 3.1.5.1 requires.that the accumulator. pressure be checked every 7 days to ensure adequate accumulator pressure. exists.
l
.to provide sufficient scram force. The' primary indicator of j
accumulator OPERABILITY is the accumulator pressure... A l minimum accumulator pressure..is specified, below which the capability of the' accumulator to perform its intended function becomes degraded and the accumulator is' considered
.inoperabl. The minimum accumulator ~ pressure of 130 Kg/c g is well. below the expected pressure of.
i 150 Kg/c g (Ref._.2). - Declaring the accumulator inoperable when the minimum pressure is.not maintained ensures that-
'l significant degradation in scram' times does not: occur. The g
- g.. f.'f 7 day frequency has been shown to be. acceptable through operating experience and takes -into account othee J
indications available in the control room.
l REFERENCES 1.
NEDE-24011-P-A, " General Electric Standard Application-Fuel,"~ September 1988.
2.
ABWR SSAR,_Section 4.6.1..
3.
ABWR SSAR, Section 5.2.2.
4.
ABWR SSAR, Section 15.4.1-
-l l
l ABWR TS B 3.1-4 10/21/93!
L l
SLC Systea B 3.1.7 8 3.1 REACTIVITY CONTROL SYSTEMS B 3.1.7 Standby Liquid Control (SLC) System BASES BACKGROUND The SLC System is designed to provide the capability of bringing the reactor, at any time in a fuel cycle, from full power and minimum control rod inventory (which is at the peak of the xenon transient) to a subcritical condition with the reactor in the most reactive xenon free state without taking credit for control rod movement. The SLC System satisfies the requirements of 10 CFR 50.62 (Ref.1) on anticipated transient without scram (ATWS).
The SLC System consists of a boron solution storage tank, two positive displacement pumps, two motor operated injection valves, which are provided in parallel for redundancy, and associated piping and valves used to transfer boratad water from the storage tank to the reactor pressure vessel (RPV). The borated solution is discharged through the "B" high pressure core flooder (HPCF) subsystem sparger.
APPLICABLE The SLC System is automatically initiated. The SLC System SAFETY ANALYSES is used in the event that not enough control rods can be inserted to accomplish shutdown and cooldown in the normal manner. The SLC System injects borated water into the reactor core to compensate for all of the various reactivity effects that could occur during plant operation. To meet this objective, it is necessary to inject a quantity of boron that produces a concentration of 850 ppm of natural boron in the reactor core at 20*C WIIEth'e"reiEf6@i9fE tiiE6Hiit Esis M iiil. Tc allcw for pcE~nti31~1EEE3FE~iRTim; idii5fiii~flhe rc :ter system, an additional amount cf bcron equal to 25% cf the amount cited above i: added
$liisfisk.C6hildinihyTUWisifilht'liiT6f!EliIhifshEfhlii
'Ref. 2' bf[RHRNatehlthsib6Eatedystnybndentfa.t be31070} ppmi(RefM 2)![org h ig hefj fod aMas sio fi water [p equallto; 1s.seldid!the1RHRishtitdoWrRE6olindjisthg?"Thi~EirisMtu thetsumt o fi the;ma s s t ofaat ernirH thsl RPViatinormaHwatsri 0
g
.gg
(,1 seesapanying LCO) are calculated such that the required concentration is achieved-accounting for dilutica in the RPY with normal water level and including the a:ter volume in (continued)
ABWR TS B 3.1-1 10/21/93
i SLC System B 3.1.7 i
BASES APPLICABLE the rc idual he:t removal shutdcwn eccling piping. This SAFETY ANALYSES quantity of borated solution is the amount that is above the (continued) pump suction shutoff level in the baron solution storage tank. No credit is taken for the portion of the tank volume that cannot be injected.
The SLC System satisfies the requirements of the NRC Policy Statement because operating experience and probabilistic risk assessment have generally shown it to be important to public health and safety.
J l
i LC0 The OPERABILITY of the SLC System provides backup capability for reactivity control, independent of normal reactivity j
control provisions provided by the control rods. The OPERABILITY of the SLC System is based on the conditions of the borated solution in the storage tank and the availability of a flow path to the RPV, including the OPERABILITY of the pumps and valves. Because the minimum required boron solution concentration is the same for both ATWS mitigation and cold shutdown (unlike some previous reactor designs) then if the boron solution concentration is less than the required limit, both SLC subsystems shall be declared inoperable. Two SLC subsystems are required to be OPERABLE, each containing an OPERABLE pump, a motor operated injection valve, and associated piping, valves, and instruments and controls to ensure an OPERABLE flow path.
APPLICABILITY In MODES I and 2, shutdown capability is required.
In M0rES 3 and 4, control rods are caly allcwed to be withdr:wn under Special Operation: LCC 3.10.3, "Ccatrol ned Withdrawal Hot Shutdcun," and LCO 2.10.4
" Control Red N
Withdrawal cold Shutdown " whichE6t?a5fi$GTsMih3Fisa b[D
" iHEa?thelfsiEf6ES63EisWiTERisiihish0tdisn?siiMillfs6ntroT s
$Ol@h~e re[a#11e@gl3hi((p6VidWd55ifiT6hfF6TsT6~
lig ensure t actor remains su6EFitiED In MODE 5, only a single control rod or control rod pai. :an be withdrawn from I
a core cell containing fuel assemblies.
Demonstration of adequate SDM (LC0 3.1.1, " SHUTDOWN MARGIN (SDM)") ensures that the reactor will not become critical. Therefore, the SLC System is not required to be OPERABLE during these conditions, when only a single control rod or control rod l
pair can be withdrawn.
l
]
(continued) l ABWR TS B 3.1-2 10/21/93 i
i SLC System B 3.1.7 l
BASES i
l ACTIONS A_d If the boron concentration is 'less than the required limits 1
given in Figure 3.1.7-1,'the concentration must be restored-to within limits in.72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ..For ATWS prevention / mitigation the ABWR features:
an automatic! rod -
insert (ARI) and an electrical insertion of FMCRDs, both of which utilize sensors and logic that are diverse and-independent of the reactor protection system; an ATWS recirculation pump trip _(RPT); and, automatic initiation of SLCS under ATWS conditions (Ref. 3). These features provide-the ABWR an ATWS prevention and mitigation ~ capability well-beyond previous BWRs. Because of the low probability,of an l
ATWS event, the ATWS prevention / mitigation features and the fact that SLC System capability.still-exists for vessel injection under these conditions ~ the _ allowed Completion Time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is acceptable and provides adequate time'to e
restore concentration to within, limits.,__.u_- - '-",_
.u_
c___,_.,_.
1,__
_, in 2...
,_ _,.. 2 irn
$YS$$'$ 5.ci$E'$'C$$dIU$.'$. E 5N id3N I ~ ~ ~
foF: RegiiPid"ActissAWiit'ibl% fist
~'
time f allowed! for. any combination' g i
limits or^ inoperable SLC ~subsystee(s;J
^
f,d contiguous occurrence of falling:to lf Condition A'is< ente' red'while, for<fas
@~systii D 0sh is inoperable and that'~subsyst'em it' i.
h@ gel $4ti i
d r
OPERABLE,'the LC0 may.:'a1 ready-have f
l 5
days.' ';This' situation:could lehd' g40
'4s th days,(7 days WCondition '8,^ fo116wed byi) 4W tisii l
A),' since' initial " failure 'of: th'e iC0(to'resto'4 j
~~
' ystem. Then'an SLC, subsystem could t e :found i i
S again', 'and concentration coulsbefesgrgte sith,11altM j
This;could,continueindefi,nitelyj i
1 MfiRIE51iff6sTriiin?nnihi" inii16wed%~jiQ@
ilf6issifsialliR"diiEi iiiWisikers efarhe'gi hiiltisg!siestab1fslijng he!#il
~ ut LC0fsistih fiallyls6thiiet-issfeud
~ % sf '
1 it
,,,,,---.1,A,, _
i q
(continued)
ABWR TS B 3.1-3
.10/21/93
SLC System B 3.1.7 BASES
't ACTIONS S_,1 (continued) i If one SLC System subsystem is inoperable, the inoperable subsystem must be restored to OPERABLE status within 7 days.
In this condition, the remaining OPERABLE subsystem.is.
adequate to perform the shutdown function. :However,'the overall reliability is reduced because a. single failure ~ in-the remaining OPERABLE subsystem could result in reduced SLC '
~
System shutdown capability. The 7. day Completion Time is based on the availability of an.0PERABLE-subsystem capable.
i of performing the intended SLC System function and the low -
probability of : 0 :i;r S: i: ?,::i t.t (0".",)..Or n,= r:
.t.
c
,__.u.
' '"

3";_';'A.'. "L'"""' l';T..! ' r 2"It.'! Ci!, y 5,,. 3
'**".b. m x x..y ; r,::,x ;, m, c; x x _.
,..,.,_i_ r__2,.,__':_!1?.~.id i
r _
een,_
.u.
ihuhjaiiisii(leum'41ms 411@I ~ }{@6 7Deg hiffi64 1
k+g'h limit on tie max
_f concentration esti4falinitspen
@ <f j ;
c 61 duri b:hnf s hgibcontl# pons;;A* MM-2 2
D i
l:,i the
,a!f Condition:8!'WB i t I
NI
/
D c" oncentration 'is 'outof lisitjs'[1M6
"?
t'o 'withincita'its' th'e;t. COMA
[
I I
$j[.
up:<to:7/ days., This situation t,i
'of 10idays%3 days in Condittoer
! W i
Condition By,'sincelinitial fatist9%
3 h, [f' the SLC:Systes.'i Then"concentpatics:
ili 1imits again,'and:the!Si,C;suk,
Me RP@A8t(lhisjsidientinue i H ThTsiC6iWhist'isn' T hiii~'ill'6irTY6FW
{
time:zero for twinningithe'411 y
res'lting in estab' ishing'>thedt u
LC0 was Initia11y':not' met' instead fM
($1 lifo36 iras' entered.; The:10tday'CompletiottT, 'is:an'
~Able 1 imitation on:this,, potential,tofat,1_te';'ae(t; l
15!!!JRit*l8i i
.Cd If both SLC subsystems are' inoperable, at-least one subsystem must be restored to.0PERABLE status within-8 hours. The allowed Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is-considered acceptable, given the low probability of :' 0**, Or -
(continued);
ABWR TS B 3.1-4 10/21/937
RCS Sp:cific Activity 3.4.6' 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.6 RCS Specific Activity i
LC0 3.4.6 The specific activity of the reactor coolant shall be limited to:
i i
a.
DOSE EQUIVALENT I-131 specific activity 5 7400 Bq/gm (0.2 pCi/gm); and b.
Grossspecificactivity53.7E+6/5/gm(100/5pCi/ge).
APPLICABILITY:
MODE 1, MODES 2 and 3 with any main steam line not isolated.
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME Ng oG ' 82 W(EU}$I6Ykgii%etp%I{n$$$$$f NEYpMIDTEN@@TMN" c3 A.
Reactor coolant specific activity WW2 Mih-Wi WM3
> 7400 Bq/gm
' 7f"6Yt*EEifnYdd$f"~
Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months N
(0.2 gCi/gm) and s 148,000 Bq/gm EQUIVALENT I-131.
(4.0 gCi/gm) DOSE EQUIVALENT I-131.
AND A.2 Restore DOSE' 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months l
EQUIVALENT-I-131 to within limits.
B.
Required Action and B.1 Determine DOSE Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months <
l associated Completion EQUIVALENT I-131.
~
Time of Condition A not met.
6M l
M B.2.1 Isolate all main 12_ hours-l-
steam lines.
l Reactor coolant specific activity E
i
> 148,000 Bq/gm (4.0 gCi/gm) DOSE.
EQUIVALENT I-131.
(continued)
ABWR TS 3.4-ff1 10/21/93 i
l
)
l i
RHR Shutdown Cooling System-Hot Shutdown 3.4.7 l
l l
3.4 REACTOR COOLANT SYSTEM (RCS) 1 3.4.7 Residual Heat Removal (RHR) Shutdown Cooling System-Hot Shutdown l
LCO 3.4.7 Two RHR shutdown cooling subsystems shall be OPERABLE, and, with less than 5 reactor internal pumps (RIPS) in operation, at least one RHR shutdown cooling subsystem shall be in operation.

NOTES---------------------------

1.
All RHR shutdown cooling subsystems and reactor internal qpAg,p3 pumps may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months g
per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided cac RHR hutdcun eccling D,
subsystem i; OPERABLE.
2, One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for performance of Surveillances provided one of the remaining RHR shutdown cooling subsystems is OPERABLE.
l l
APPLICABILITY:
MODE 3 with reactor steam dome pressure < 9.5 Kg/cm2g (135 psig).
l ACTIONS
___________________.._-NOTE-------------------------------------
LC0 3.0.4 is not applicable.
i i
CONDITION REQUIRED ACTION COMPLETION TIME i
A.
One or two required A.1 Initiate action to Immediately RHR shutdown cooling restore required RHR subsystems inoperable.
shutdown cooling subsystem (s) to OPERABLE status.
l AND (continued)
ABWR TS 3.4-1 10/21/93
RHR Shutdown Cooling Systca-Cold Shutd:wn f
3.4.8 i
3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.8 Residual Heat Removal (RHR) Shutdown Cooling System-Cold Shutdown-i i
LC0 3.4.8 Twehiris RHR shutdown cooling subsystems shall be OPERABLE, t
and;^*WTth less than 5 reactor internal pumps (RIPS) in i
operation, at least one RHR shutdown cooling ~ subsystem shall be in operation.
_________...__...______...--NOTES---------------------------
1.
ALL RHR shutdown cooling subsystems and reactor internal L
pft'Dg pumps may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months c, 4-per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided on : b:y:t = i: CPEPfSLE.
l 2.
One RHR shutdown cooling subsystem may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the performance of Surveillancesr
,7 pd pecvided :n: Of the re::inf,
..? :hutd;wn :: cling D
-+-
sub:y:te:: is OPERABLE :nd
+o d4en.
3M['0hERHR?slidfd6EEE66TT6F?iGdsysf5sIssEli4Mii6FEFA@
af te M301hoia r sjftomli n i t] a]j[en} ryli nto[ NODE 41from MDE3r t
s APPLICABILITY:
MODE 4.
^
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.
One or IEIFs required A.1 Verify an' alternate 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months RHR shufd6sn cooling method of decay heat subsystems inoperable.
removal 'is available AND for each required inoperable RHR Once per shutdown cooling 24_ hours.
subsystem.
thereafter-(continued)
ABWR TS 3.4 1 10/21/93 i
4
(
.RCS Sp:cific Activity-i
~B 3.4.6-
.1 BASES EH6f0T l
(fiLC0} EWE?RUDNSEAIcI3idis MiiThis$ stspiilohi.
i ACTIONS o
~
E511ssMs1N00E(S}ynhilus ;
~~
7
,(hejACTIONSjaaygeventHE d
XCeRINNd5II Ik bgo&
1prporateE.
hs d V i Q ls $jp....
_.Xcurs chi % ~~l y_
ProbhMTMK
fi _
O$
thliiT151 isMch@tdjM606iil"dinif)th~~..
i Ulfema i
~
l t
A.1 and A.2 When the reactor coolant specific activity exceeds the LC0 I
DOSE EQUIVALENT I-131 limit, but.is s~148,000 Bq/gm (4.0 gCi/ a),: samples must be analyze.d for DOSE EQUIVALENT I-131 l
at least once every 4_ hours.
In addition, the specific.
F activity must be restored to the LCO limit within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .-
The Completion Time of once'every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months isibased on the
.l time needed to take and analyze a sample.H The 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ' Completion Time to restore the activity level provides a.
reasonable time for temporary coolant activity increases i
(iodine spikes or crud bursts) to be cleaned up'with the-normal processing systems.
B.1. B.2.1. B.2.2.1. and B.2.2.2 If.the DOSE EQUIVALENT I-131 cannot be restored to s ~7400 Bq/gm (0.2 pCi/gm) within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or if at any time it~is;.
> 148,009 Bq/gm (4.0 yCi/gm), it must'he determined at-least.
every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and all the main steam lines must be: isolated t
within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ' Isolating the main steam: lines precludes i
the possibility of releasing radicactive material to the.
environment in an amount that is more than.a small fraction' of the requirements of 10 CFR 100 during a postulated MSLB j
accident.
f i
Al!.ernately, the plant can be brought to MODE 3 within.
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36; hours. This option is-provided for those instances when isolation of main steam lines is not desired (e.g., due to the decay heat loads).
In MODE 4, the requirements of the_ LCO are no longer i
appli:able.
l (continued)-
f ABWR TS B 3.4-3 10/21/93.
l 1
RHR Shutdown Cooling Systcm-Hot Shutdown B 3.4.7-BASES LC0 cooling subsystem consists of one OPERABLE RHR pump, a heat (continued) exchanger, and the associate piping and valves. Each shutdown cooling subsystem is considered OPERABLE if it can j
be manually aligned (remote 3r local) in the shutdown l
cooling mode for removal of decay heat.
In MODE 3, one RHR i
shutdown cooling subsystem can provide the required cooling, but two subsystems are required to be OPERABLE to provide i
redundancy. Operation of one subsystem can maintain or reduce the reactor coolant temperature as required.-
However, to ensure adequate core flow to allow for accurate average reactor coolant temperature monitoring, nearly continuous operation is required.
i l
Note 1 permits all RHR shutdown cooling subsystems and reactor internal pumps to be shut down for a period of
@4p 03 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period, provided onc eb:y;te i:
OPER?SLE. Note 2 allows one RHR shutdown cooling subsystem d
to be inoperable for up t0 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; for performance of surveillance tests ;rovided one of the rc::ining RS i
shutdcun eccling ;u::ystem; i: OPEPf3LE. These tests may be on the affected RHR System or on some other plant ~ system or component that necessitates placing the RHR system in an _
i inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of redundancy.
APPLICABILITY In H0 DES 1 and 2, and in MODE 3 with reactor steam dome pressure above the RHR cut in permissive pressure, this LC0 l
1s not applicable. Operation of the RHR System in the shutdown cooling mode is not allowed above.this pressure because the RCS pressure may exceed the design pressure of the shutdown cooling piping. Decay heat removal at-reactor pressures above the RHR cut in permissive pressure is typically accomplished by condensing the steam in the main condenser. Additionally, in MODE 2 below this pressure, the OPERABILITY requirements for the Emergency Core Cooling Systems (ECCS) (LCO 3.5.1, "ECCS-Operating") do not allow placing the low pressure RHR shutdown cooling subsystem into operation.
(continued) 3 ABWR TS B 3.4-2 10/21/93 l
l RHR Shutdown Cooling System-Cold Shutdown l
B 3.4.8 1
BASES LC0 cooling subsystem' consists of one OPERABLE RHR pump,.a heat (continued)
- exchanger, and the associated piping ~~and valves-Each shutdown cooling subsystem is considered OPERABLE if it can be manually aligned'(remote or-local) in the shutdown cooling mode for removal of. decty heat.
In MODE 4,.one RHR l
shutdown cooling subsystem can provide the required cooling,.
but-two subsystems'are required-to be 0PERABLE to provide redundancy.
Operation of one subsystem can maintain and-i reduce the reactor coolant temperature as required.-
However, to ensure adequate core-flow.to allow for accurate' j
average reactor coolant temperature monitoring, nearly '
j continuous operation is required.
l Note 1 permits'all RHR shutdown cooling -subsystemsiand ' RIPS q
to be shut down for a period of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in an 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period,
<l Op3
--pr-evided := cubrystem i: CPERACLE Note 2 a"ows.one RHR shutdown cooling subsystem to be inoperable
,, up to Ch O 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for the performance of surveillance testsgevided i
cr.: cf-thc rc=i-ing P" chutdewa meling rubrycter: is -
OPERACLE. These tests may be on the affected RHR-Eystem or!
l on some other plant. system'or component that necessitates.
placing the RHR System in an. inoperable status during the.
performance.
This is ' permitted because the core heat-l generation can be low enough and the.heatup-rate slow enough I
to allow some changes to the RHR subsystems,or other-
[
operations requiring RHR flow interruption and loss ofl l
redundancy.
l
!.j
-i t
APPLICABILITY In MODES 1 and 2, and in MODE 3 with reactor steam dome pressure above the RHR cut in permissive.pressuref this;LC0 l
is not applicable. Operation of-the RHR System.in the.
I shutdown cooling mode is not allowed above this pressure -
because the RCS pressure may exceed the design pressure of.
the shutdown cooling piping. Decay heat removal at reactor pressures above the RHR cut in permissive pressure is-J typically accomplished by condensing the steam in the main
.I condenser. Additionally, in MODE 2 below this; pressure, the j
OPERABILITY requirements for-the: Emergency Core Cooling
-l Systems (ECCS) (LC0 3.5.1, "ECCS-Operating") do not allow placing the low pressure RHR shutdown cooling' subsystem into operation.
APPLICABILITY In MODE 4, the RHR System may be operated in=the' shutdown
)
1 (continued)'
i ABWR STS B 3.4-36 10/21/93 l
l Primary Containment Hydrogen Recombiners 3.6.3.1 l
3.6 CONTAINMENT SYSTEMS 3.6.3.1 Primary Containment Hydrogen Recombiners LC0 3.6.3.1 Two primary containment hydrogen recombiners shall be OPERABLE.
APPLICABILITY:
MODES I and 2.
ACTIONS l
CONDITION REQUIRED ACTION COMPLETION TIME A. One primary A.1

NOTE---------

containment hydrogen LC0 3.0.4 is not i
recombiner inoperable.
applicable.
Restore primary 30 days l
containment hydrogen i
l recombiner to OPERABLE status.
4 8.
Two primary B.1 Verify by I hour containment hydrogen administrative means recombiners that the hydrogen AND inoperable.
control function is g g-c6 I
maintained.
Once per g
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND B.2 Restore one primary containment hydrogen 7 days recombiner to OPERABLE status.
l C.
Required Action and C.1 Be in MODE 3.
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met.
l ABWR TS 3.6-29 10/21/93 l
CRHA AC Systeo i
B 3.7.5 k
i BASES ACTIONS E.1. E.2. and E.3 (continued)..
a potential for releasing radioactivity that might require i
isolation of the control room. This places the unit in a j
condition that minimizes risk.
I If applicable, CORE ALTERATIONS and handling of irradiated fuel in the primary or secondary containment must be suspended immediately. Suspension of these activities shall' l
not preclude completion of movement of a component to a safe l
l position. Also, if applicable,. actions must be initiated immediately to suspend OPDRVs to minimize the probability of a vessel draindown and subsequent potential for fission product release. Actions must continue until the OPDRVs are suspended.
SURVEILLANCE SR 3.7.5.1 REQUIREMENTS This SR verifies that the heat removal capabilit of the s.ctem is sufficient to remove the NCAEihsif!T6i iTsiGiiisaTjj d oG '2 Y
"isififf~d'Eildilifion.fisslyisil The SR consisfilitT~diis5T fro ~6~^6f"
' hg'in The 18 month Frequency is i
c4 appropriate since significant degradation of the CRHA AC System is not expected over this time period.
l REFERENCES 1.
ABWR SSAR, Section 6.4.
2.
AT4R SSAR, Section 9.4.1.
l l
9 ABWR TS B 3.7-f 10/21/93
)
-l l
RHR-High Wat:r Level j
i l
3.9.7 I
l I
l 3.9 REFUELING OPERATIONS i
l 3.9.7 Residual Heat Removal (RHR)-High Water Level LC0 3.9.7 One RHR shutdown cooling subsystem shall be OPERABLE and in operation.
i
_______..---N0TE-----------------------r----
The required RHR shutdown cooling subsystem may be'remov'ed-i from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per.8-hour period.
APPLICABILITY:
!{ '
l p^
l6
[
{lf_
RPY~ 1i5GE':~~
6tJ A,- 03 1
ACTIONS CONDITION REQUIRED ACTION
-COMPLETION TIME A.
Required RHR shutdown A.1 Verify an alternate-I hour cooling subsystem method of decay heat -
inoperable.
removal is available.
Ag Once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s-thereafter l
B.
Required Action and B.1 Suspend. loading Immediat'ely.
[
associated Completion irradiated fuel l
Time of Condition A assemblies into the-not met.
RPV.
AND l
B.2 Initiate action to
'Immediately restore primary or l
secondary' containment to OPERABLE status.
A@
(continued) l ABWR TS 3.9-1 10/21/93 L
RHR-Low Water Level 3.9.8 3.9 REFUELING OPERATIONS 3.9.8 Residual Heat Removal (RHR)-Low Water Level LCO 3.9.8 Two RHR shutdown cooling subsystems shall be OPERABLE, and one RHR shutdown cooling subsystem shall be in operation.
............................N0TE-------------
2-------------
The required operating chutdown cooling subsystem may be removed from operation fer up to 2 hears per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period.
APPLICABILITY:
MODE 5 with 1FFsdiitidifsi1QE%ihi?Fissf6F?FfissiaiFiiViss'st
'(RPU?i6fMfiheliGF TeVelT7^.'o~iii~ibsue^^thsT6F'5f~'the RPV flange.
h u)& 43 64 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME' A.
One or two RHR A.1 Verify an alternate 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months shutdown cooling method of decay heat.
subsystems inoperable.
removal is available A!iQ for each inoperable RHR shutdown cooling Once per subsystem.
24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter BM_Riddffsd7Asti6nishd B._il Initiate action to Immediately
,g g
TimeY6flConditionW^
secondary containment h6 tim 6tf~~~'~~~
to OPERABLE status.
ANQ h8 Bi g ] Initiate action to Immediately gg restore one standby gas treatment subsystem to OPERABLE status.
AND (continued) l
[
ABWR TS 3.9-1 10/21/93 4
l
.J I
i RHR-Low Water Level 3.9.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1
B.
(continued)
Bi3l Initiate action to Immediately restore one secondary 3
containment isolation valve and associated instrumentation to OPERABLE status in each associated penetration flow path not isolated.
C.
No RHR shutdown C.1 Establish reactor I hour from
~
coolant circulation discovery of no cooling subsystem in operation.
by an alternate reactor coolant method.
circulation AND 4
C.2 Monitor reactor Once per hour coolant temperature.
-f SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
]
SR 3.9.8.1 Verify one RHR shutdown cooling subsystem 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is operating and circulating reactor coolant.
i i
i ABWR TS 3.9-2 10/21/93
RHR.- High Water Level B 3.9.7 BASES LC0 An OPERABLE RHR shutdown cooling subsystem consists of an (continued)
RHR pump, a heat exchanger, valves, piping, instruments,-and controls to ensure an OPERABLE flow pa:h.
Additionally, each RHR shutdown cooling subsystem.is considered OPERABLE if it can be manually aligned (ramote or i
local) in the shutdown cooling mode for removal of decay l
heat. Operation (either continuous'or intermittent)'of one l
subsystem can maintain and reduce the reactor coolant-temperature as required. However, to ensure adequate. core flow to allow for accurate average reactor coolant i
temperature monitoring, nearly continuous operation -is '
l required. A Note is provided to allow a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exception to shut down the operating subsystem every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
l APPLICABILITY One RHR shutdown cooling subsystem is required to be OPERABLE in MODE 5 iil thliVFadistEffislM6?tFsIFsibi jifi_sssF.ETVisss1N.68%ith the~ Wit'sF"1EVs1T770 tis 6iT@b the i
g p
n o rovide decay heat removal. RHR System requirements in other MODES are covered by LCOs in pf,p3 Section 3.4, Reactor Coolant System-(RCS); Section 3.5, l
C9 Ecergency Core Cooling Systems (ECCS); and Section 3.6, Containment Systems. RHR System requirements in MODE 5 T.FFidlitsdi.fus1?lnfihi?F#ictsFIFFissufiWissilysWdiWTih, with the gg g ~C 770~Eb^5s~16s~RPC fliiiii,~iFi' g IVid'IE l
LCO 3.9.8, " Residual Heat Removal (RHR) - Low Water. Level."
i ACTIONS
.A_J j
With no RHR shutdown cooling subsystem OPERABLE, an alternate method of decay heat removal must be established within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . In this conditioni, the volume of water above the RPV flange provides adequate capability to remove decay l
heat from the reactor core. However, the overall reliability l
is reduced because loss of water level could result in i
reduced decay heat removal capability. The I hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal l
capabilities. Furthermore, verification of the functional availability of these alternate method (s) must be l
{
(continued)
ABWR TS B 3.9-23 10/0/93
n RHR - Low Water Level B 3.9.8 BASES LC0 Additionally, each RHR shutdown cooling subsystem is (continued) considered OPERABLE if it can be manually aligned (remote or local) in the shutdown cooling mode for removal of decay heat. Operation (either continuous or intermittent) of one subsystem can maintain and reduce the reactor coolant temperature as required. However, to ensure adequate core flow to allow fer accurate average reactor coolant temperature monitoring, continuous operation is required. A Note is provided to allow a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months exception to shut down the operating subsystem every 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
i APPLICABILITY Two RHR shutdown cooling subsystems are required to be OPERABLE in MODE 5,@WithWFeadlitedif0ellihiths'Fsiit6Ei t pfesidfeNssss17shd top of ths'RPV~ flange, to provide decay heat removal. RHR pg.M System requirements in other MODES are covered by LCOs in 6Y Section 3.4, Reactor Coolant System (RCS); Section 3.5, Emergency Core Cooling Systems (ECCS); and Section 3.6, Containment Systems. RHR System re,quirements in MODE 5, with IFridl.itsd)?f6elEthithsTFFictdEJFiss0fiWisiiWihdfWfth the g
WE C0 YibBiie't hi~ R PV " flih5E,"^iFF'iiVeirTi' LC0 3.9.7, " Residual Heat Removal (RHR)- High Water Level."
ACTIONS A.1 With one of the two required RHR shutdown cooling subsystems inoperable, the remaining subsystem is capable of providing the required decay heat removal. However, the overall reliability is reduced. Therefore an alternate method of decay heat removal must be provided (such as the third RHR shutdown cooling subsystem). With both required RHR shutdown cooling subsystems inoperable, an alternate method of decay heat removal must be provided in addition to that provided i
for the initial RHR shutdown cooling subsystem inoperability. This re-establishes backup decay heat removal capabilities, similar to the requirements of the LCO. The I hour Completion Time is based on the decay heat removal function and the probability of a loss of the available decay heat removal capabilities. Furthermore, verification of the functional availability of these alternate method (s)
(continued)
ABWR TS B 3.9-27 10/21/93
RHR -~Lcw Water Lcvsl i
B 3.9.8 BASES ACTIONS A.1 (continued)
I must be reconfirnad every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. This will ensure continued heat removal capability.
Alternate decay heat removal. methods are available to.the i
operators for review and preplanning in the unit's Operating l
Procedures. For example, in addition to. the third RHR-shutdown cooling loop, this may include the use of the Reactor Water Cleanup System, operating with the I
regenerative heat exchanger bypassed. The method used to remove decay heat should be the most prudent choice based on unit conditions.
pO3 BY W T 2? T 3? T M *TaTd T f l
V/V>
If no RHR shutdown cooling subsystem is in operation, an--
alternate method of coolant circulation is required to be established within I hour. The Completion Time.is modified such that the I hour is applicable separately for each occurrence involving a loss of coolant circulation.
During the period when the reactor coolant.is' being l
circulated by an alternate method (other than by the required RHR Shutdown Cooling System), the reactor coolant temperature and level must be periodically monitored to -
ensure proper function of the alternate method. The.once per hour Completion Time is deemed appropriate.
If at least one RHR subsystem is not restored to OPERABLE
\\
status immediately, additional actions are required to minimize any potential fission product release to the.
environment. This includes initiating immediate action to restore the following to OPERABLE status: secondary.
j containment, one standby gas treatment subsystem, and one i
secondary containment isolation valve. and associated i
instrumentation in each associated penetration not isolated.
~
This may be performed as an administrative check, by examining logs or other information to determine whether the l
components are out of service for maintenance or other reasons. It is not necessary to perform the surveillances.
l needed to demonstrate the OPERABILITY of the components. If, I
however, any required component is inoperable, then it must I
be restored to OPERABLE status. In this case, the (continued)
ABWR TS B 3.9-28 10/21/93 i
RHR - Low Watar-Level l
B 3.9.8 BASES ACTIONS BIFBI2NBl3?c?1?7iEFcif' (^s'6sHWiisd)':
' "~^'~~~~~
(continued) l surveillance may need to be performed to restore the component to OPERABLE status. Actions must continue'until all required components are OPERABLE.
i SURVEILLANCE i
REQUIREMENTS SR 3.9.8.1 This Surveillance demonstrates that one RHR subsystem is in.
l operation and circulating reactor coolant. Tne required flow-l rate is determined by the flow rate necessary to provide sufficient decay heat removal capability. The' Frequency of l
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient in view of other visual and audible l
indications available to the operator for monitoring _the RHR-l subsystem in the control room.
I l
REFERENCES None.
i I
.I l-ABWR TS B 3.9 29 10/21/93 i
I
i SDM Test-Refueling 3.10.8 l
3.10 SPECIAL OPERATIONS 3.10.8 SHUTDOWN MARGIN (SDM) Test-Refueling i
LC0 3.10.8 The reactor mode switch position specified in Table 1.1-1 for MODE 5 may be changed to include the startup/ hot standby position, and operation considered not to be in MODE 2, to l
allow SDM testing, provided the following requirements are met:
6g0I b iT(('[C6?fRI?5G$$5[DII$5hT5%f6IfWillE6fif@jfdh1%f3 t
"^"
FF$ifeinshts}fshif0Et{6hs12Jsjldgjdj 4
LablEMDlEh i
b.
1.
LCO 3.3.5.1, " Control Rod Block Instrumentation,"
MODE 2 requirements for Function 1.b of-Table 3.3.5.1-1, i
OR l
2.
Conformance to'the approved control rod sequence for the SDM test is verified by.a second licensed' operator or other qualified member of the technical l
staff; c.
Each withdrawn ' control rod shall be coupled to the.
associated CRD; d.
All control rod withdrawals [during out of sequence control rod moves] shall be made in notch out mode; and e.
No other CORE ALTERATIONS are in progress.
APPLICABILITY:
MODE 5 with the reactor mode switch in startup/ hot standby l
position.
l l
l 1
ABWR TS 3.10-20 10/21/93
I l
SDM Test-Refueling 3.10.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i
A.
One or more of the A.1 Place the reactor-Immediately above requirements not mode switch in the met, for reasons other shutdown or refuel than Condition B.
position.
E p\\
r M
B.
One control rod not B.1 Declare the affected Immediately i
4 4
[f coupled to its control rod associated CRD.
inoperable.
r i
t SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l
t 0\\-
[0 SRMM018 1sC[Piff6iWMii[applj dblij}Riiffr3COM1311}M
$266RfnillB_
9 unctions;2(alandj23J thelapdici@3 1
-r SRs SR 3.10.8.2

NOTE--------------------

NotrequiredtobemetifSR3.10.8.y l
satisfied.
Perform the applicable SRs for LCO 3.3.5.1, According to Function 1.b.
the applicable SRs l
(continued)
I ABWR TS 3.10-21 10/21/93
SDM Test-Refu)11ng 3.10.8 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.10.8.3

NOTE--------------------

~
Not required to be met if SR 3.10.8.R satisfied.
Verify movement of control rods is in During control compliance with the approved control rod rod movement sequence for the SDM test by a second licensed operator or other qualified member of the technical staff.
SR 3.10.8.4 Verify no other CORE ALTERATIONS are in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months progress.
SR 3.10.8.5 Verify each withdrawn control rod does not Each time the go to the withdrawn overtravel position.
control rod is withdrawn to
" full out" position A!!Q Prior to satisfying LCO 3.10.8.*d requirement' after work on control rod or CRD System that could affect coupling ABWR TS 3.10-22 10/21/93
i l
Shutdown Margin (SDM) Test-Refueling B 3.10.8 BASES APPLICABLE RWE analyses assume that the reactor operator follows SAFETY ANALYSES prescribed withdrawal sequences.
For SDM tests performed (continued) within these defined sequences, the analyses of References I and 2 are applicable. However, for some sequences developed for the SDM testing, the control rod patterns assumed in the safety analyses of References 1 and 2 may not be met.
Therefore, special RWE analyses, performed in accordance with an NRC approved methodology, are required to l
demonstrate that the SDM test sequence will not result in l
unacceptable consequences should a RWE occur during the testing.
For the purpose of this test, protection provided by the normally required MODE 5 applicable LCOs, in addition l
to the requirements of this LCO, will maintain normal test operations as well as postulated accidents within the bounds of the appropriate safety analyses (Refs. B3.10.8-1 and B3.10.8-2).
In addition to the added requirements for the RWM, SRNM, APRM, and control rod coupling, the notch out mode is specified for out of sequence withdrawals.
Requiring the notch out mode limits withdrawal steps to a l
single notch, which limits inserted reactivity, and allows adequate monitoring of changes in neutron flux, which may l
occur during the test.
As described in LC0 3.0.7, compliance with Special Operations LCOs is optional, and therefore, no criteria of i
the NRC Policy Statement apply.
Special Operations LCOs provide flexibility to perform certain operations by appropriately modifying requirements of other LCOs. A discussion of the criteria satisfied for the other LCOs is provided in their respective Bases.
LC0 As described in LCO 3.0.7, compliance with this Special Operations LC0 is optional.
SDM tests may be performed while in MODE 2, in accordance with Table 1.1-1, without meeting this Special Operations LCO or its ACTIONS.
For SDM tests performed while in MODE 5, additional requirements must be met to ensure that adequate protection against potential reactivity excursions is available.
Because multiple control rods will be withdrawn and the reactor will potentially become critical, RPCMODEI2?FEq'GWiisshtiifu 6 # 01A Functi6 nil 2?aiand E 2WuGTib1M 333hWEmustThehnfoFdd
c i and ~th6'5i$Fivid ~E6nt'F6Tf6d~Wi th'd FissT~isqss~nn"iiiGst"b6~"
enforced by the RWM (LC0 3.3.5.1, Function Ib, MODE 2), or must be verified by a second licensed operator or other (continued)
ABWR TS B 3.10-34 10/21/93
ShutdIwn Margin (SDM) Test-Refueling B 3.10.8 BASES LC0 (continued) qualified member of the technical staff. To provide additional protection against an inadvertent criticality, control rod withdrawals that do not conform to the ganged withdrawal sequence restrictions specified in LCO 3.1.6,
" Rod Pattern Control" (i.e., out of sequence control rod withdrawals) must be made in the notched withdrawal mode to minimize the potential reactivity insertion associated with each movement. Coupling integrity of withdrawn control rods is required to minimize the probability of a RWE and ensure proper functioning of the withdrawn control rods, if they are required to scram.
Because the reactor vessel head may be removed during these tests, no other CORE ALTERATIONS may be in progress. This Special Operations LC0 then allows changing the Table 1.1-1 reactor mode switch position requirements to include the startup/ hot standby position, such that the SDM tests may be performed while in MODE 5.
APPLICABILITY These SDM test Special Operations requirements are only applicable if the SDM tests are to be performed while in MODE 5 with the reactor vessel head removed or the head bolts not fully tensioned. Additional requirements during these tests to enforce control rod withdrawal sequences and restrict other CORE ALTERATIONS provide protection against potential reactivity excursions. Operations in all other MODES are unaffected by this LCO.
ACTIONS A.1 With one or more of the requirements of this LC0 not met, the testing should be immediately stopped by placing the reactor mode switch in the shutdown or refuel position.
This results in a condition that is consistent with the requirements for MODE 5 where the provisions of this Special Operations LC0 are no longer required.
1 El With the requirements ~of thi's' LCO n~o't met,~ th'6?iffsetsa "
Q A-CM control rod shall be declared inop' rable. This results' E~a e
gg conditionthatisconsistentwith_thegeguire,ments,for_ MODE ~
(continued)
ABWR TS B 3.10-35 10/21/93
l Shutdown Margin (SDM) Test-Refueling B 3.10.8 BASES ACTIONS
[(poh{ihE{dJGf((1Rn9Estequigdj$3 M (fR @sJp M Q jhji6fi @i ESpjjlpl[DhH 2
I SURVEILLANCE SRM3?t018.71 REQUIREMENTS Piif6HiiisWEsI6HfhiWpMETTEiSTETsRiW tDMMTQ7 pold Fuiictionis2?sinsd?2fdWillishineitnt theWeictis 6Mpatjgs!tpfflyMjj(j{ofjahjjl({jj 1g c, i.
SR 3.10.8.2 and SR 3.10.8.'3 The control rod withdrawal sequences during the SDM tests may be enforced by the RWM (LC0 3.3.5.1, Function Ib, MODE 2 requirements) or by a second licensed operator or other qualified member of the technical staff. As noted, either the applicable SRs for the RWM (LCO 3.3.5.1) must be satisfied according to the applicable' Frequencies (SR 3.10.8.1 and SR 3.10.8.2), or the proper movement of control rods must be verified. This latter verification (i.e., SR 3.10.8.2) must be performed during control rod movement to prevent deviations from the specified sequence.
These surveillances provide adequate assurance that the specified test sequence is being followed..
SR 3.10.8.4 Periodic verification of the administrative controls established by this LCO will ensure that the reactor is
]
operated within the bounds of the safety analysis. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is intended to provide appropriate assurance that each operating shift is aware of and verifies compliance with these Special Operations LCO requirements.
SR 3.10.8.5 Coupling verification is performed to ensure the control rod is connected to the control rod drive mechanism and will perform its intended function when necessary. The-verification is required to be performed any time a control rod is withdrawn to the' " full out" notch position or prior (continued)
ABWR TS B 3.10-36 10/21/93
i Shutdown Margin (SDM)_ Test-Refueling B 3.10.8 BASES SURVEILLANCE skFis?id?875 (continued)
REQUIREMENTS to declaring the control rod OPERABLE after work on the l
control rod or CRD System that could affect coupling. Tht; Frequency is' acceptable, considering the low probability that a control rod will become uncoupled when it is not being moved as well as operating experience related to uncoupling events.
l REFERENCES 1.
NEDE-24011-P-A-US, General Electric Standard Application for Reactor Fuel, Supplement For United-States (as amended).
2.
ABWR SSAR, Section 15.4.1.
f I
h 4
b t
ABWR TS B 3.10-37
.10/21/93-
Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum k-infinity of 1.35 in the normal reactor core configuration at cold conditions; b.
k 'ludes an allowance for uncertainties as described ins 0.95 if fully floo iNc l
Section 9.1 of the SSAR; t
CDb,;th c.
A nominal fuel :::::bly center te center stcr:ge : pacing of [
] ::, within : neutron poi:en ::teri:1 between Y
stcr:ge :p;cc:, in the high density :ter:ge r:ck: in the
pent fue': ster:ge p;;1.
4.3.1.2 The new fuel storage racks are designed and shall be i
maintained with:
a.
Fuel assemblies having a maximum k-infinity of 1.35 in the normal reactor core configuration at 20 *C; i
l b.
k 5 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.1 of the SSAR; c.
k,,, s 0.98 if moderated by aqueous fo&m, which includes an allowance for uncertainties as described in -
Section 9.1 of the'SSAR; and d.
A nominal [ ] cm center to center distance between fuel assemblies placed in storage racks.
4.3.2 Drainaae The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below 3.1 m (10 ft) above the top of the active fuel.
(continued)
ABWR TS 4.0-2 10/21/93 1
I l-.
%

ML20058L922

Text

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