ML20134E034

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Informs That Staff Unable to Review Application to Adopt STSs for Entire Design Features Section Before Next ANO-2 Refueling Outage & Recommends That Licensee Submit Request for Specific Changes Needed in Separate Application
ML20134E034
Person / Time
Site: Arkansas Nuclear 
Issue date: 10/28/1996
From: Salehi K
NRC (Affiliation Not Assigned)
To: Hutchinson C
ENTERGY OPERATIONS, INC.
Shared Package
ML20134E038 List:
References
TAC-M96480, NUDOCS 9610310192
Download: ML20134E034 (38)
v • d • e


Text

_ _ _

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l UNITED STATES y

g NUCLEAR REGULATORY COMMIS810N j

waswimorou, o.c. men.ean October 28, 1996 l

l Mr. C. Randy Hutchinson i

Vice President. Operations ANO j

Entergy Operations, Inc.

1448 S. R. 333 l

Russellville, AR 72801

SUBJECT:

CONVERSION OF THE DESIGN FEATURES SECTION OF THE TECHNICAL SPECIFICATIONS TO THE IMPROVED STANDARD TECHNICAL SPECIFICAT!0NS (ISTS), ARKANSAS NUCLEAR ONE, UNIT 2 (ANO-2) (TAC N0. M96480)

Dear Mr. Hutchinson:

By letter dated August 23, 1996, Entergy Operations, Inc. (E01), submitted the i

subject application.

In that letter E01 requested that the Nuclear Regulatory i

Connission (NRC) complete its review prior to the next ANO-2 refueling outage (2R12), which is currently scheduled to begin on April 11, 1997.

The NRC staff perfonned an initial review of the application. The application j

adopts the improved standard technical specifications (ISTS) for the entire Design Features section. To perform a complete review of the application, it j

would require a significant staff effort as each change would need to be i

individually evaluated.

In addition, the staff noted differences between the j

application and the ISTS which would further lengthen the review process.

Due to NRC's priority of reviewing full conversion a>plications being ahead of partial conversion applications, the staff will not >e able to meet E01's requested review schedule. Accordingly, the staff recommends that you identify the specific changes that are needed to support refueling outage 2R12, and request those changes in a separate application.

Sincerely, Kombiz Salehi, Actinti Project Manager Project Directorate :V-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Docket No. 50-368 cc: See next page ggo3/0/7Z O

NUMBERS.SGM SGML Technical Specificatior.s - Format for Headings Chapter /Section Numbering and Names Section 3.0 should NOT have same number as Chanter 3.0 Alt 1:

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Spec Name Where Used 2.0 Safety...litits g )?idd Limi TOC and H1 KWie ons 2.1 B,s~~~~~~'~~

TOC and H2 2.2 5.fL^ Violations TOC and H2 3.0

^?"siiul!SRs TOC Only EA UAsqufriiiiiiiti TOC and H1

( 3~.'0)

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LC0~Ap511cabiliti' TOC and H2 (3.0) 8; A'.2 SR Applicatility TOC and H2 3.1 Reactivity Control TOC and H1 Systems 3.1.1 Shutdown Margin TOC and H2 Alt 2:

(Chapter numbers as a single digit, 3.0 is a Section number)

Chapter Section Spec Name Where Used 1

Use and Application TOC and H1 1.1 Definitions TOC and H2 1.2 Logical Connectors TOC and H2 1.3 Completion Times TOC and H2 1.4 Frequency TOC and H2 Safety. Limits.l$Lfand TOC and HI 2

SL; Violations 2.1 SLs TOC and H2 2.2 S.L' Violations TOC and H2 LCOsiand?SRs_..

TOC Only 3

Seneral!Requirem_ents TOC and H1

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3;0 i

$10.1 LCO Applicability TOC and H2 3;0;2 SR Applicatility TOC and H2 3.1 Reactivity Control TOC and H1 Systems 3.1.1 Shutdown Margin TOC and H2

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Design Features TOC and H1 4.1 Site Location TOC and H2 i

4.2 Reactor Core TOC and H2 4.3 Fuel Storage TOC and H2 4

l Administrative Controls TOC and HI 5.1 Responsibility TOC and H2 5.2 Organization TOC and H2 5.3 Unit Staff Qualifications TOC and H2 5.4 Procedures TOC and H2 5.5 Programs and Manuals TOC and H2 5.6 Reporting Requirements TOC and H2 5.7 High Radiation Area TOC and H2 TYPICAL PAGE i

Definitions 1.1

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Use and Application jyCliipth]litli's7hNiisifill3ajilfaljiikliifirs t I

1.1 Definitions 1

TYPICAL PAGE SLs i

2.1

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SafetyLimits(SLs)[ssd?SQyibistidhi 2.1 SC$

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I TYPICAL PAGE x

LC0 Applicability sj0:1 M

eisiEillMfNiisti Ki((LimitingConditionforOperation(LCO) Applicability LCO 3.0.1 LCOs shall be met during the MODES or other specified

TYPICAL PAGE SR Applicabi.lity bp22 li!

E55!51MIMH

$g2 Surveillance Requirement (SR) Applicability SR 3.0.1 SRs shall be met during the MODES or other specified 1

1 TYPICAL PAGE SDM 3.1.1 l

3.1 Reactivity Control Systems

LLO 3.1.1 The SDM shall be [ greater than or equal to the limit specified in the COLR. The minimum limit shall be] 2 [1.0]% Ak/k.

TYPICAL PAGE S. it. (.16 cation

[

Design Features

[{l Site Location

[ Text description of site location.)

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TYPICAL PAGE grgarcsa 112 1

l Design Features

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Reactor Core 4.2.1 Fuel Assemblies j

The reactor shall contain [177] fuel assemblies.

Each assembly

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1 i

i l

TYPICAL PAGE Responsibility 5.1 4

5 Administrative Controls 5.1 Responsibility j

5.1.1 The [ Plant Superintendent) shall be responsible for overall unit t

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1 3

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i W Q141865 TOC. MOD IBS%Dt!!*Aletter caHIgli TABLE OF CONTENTS l$

Use and Application 7151 i

1.1 Definitior.:.

tifl i

1.2 Logical Connectors..................

I 1.3 Compl e t i on T i me s...................

1 1.4 rrequency

' 1 2@1 Safety Limits (SLs)[jid M M 61 M @

M j

2.

SLs 1

2.2 SL Violations

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$5filiptist1Y1aWFf6F"5sFitiMt20Fiiinll$nt11isiBii FatequirementiffM)UMRMtphenyMssh s

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$;0T^Z"8eneral ? Requiriments TlRRbbish6M5@en@L'hj31031:1 d

3.0.1 Limitins Conditions 'f6F^0piFitio6~Mbli&bil ft'i'.'~~.~.'~". ~. 810:1-1 3.0.2 Surveillance Requirements Applicability........

3_:0_;2-1 3

4 3.1 Reactivity Control Systems.............

3!!!121 3.1.1 Shutdown Margin (SDM) 8 ?l 21 - I i

3.1.2 Reactivity Balance...............

3;}]2-1 4l Design Features

..................... Bilfl 4 ~.'1 Site Location 41151 i

4.2 Reactor O re....................

4?291 l

4.3 Fuel Sto rage..................... Q-1 1

SE Administrative Controls 551 571 Responsibility.................... $1141 5.2 Organization....................

5:241 5.3 Unit Staff Qualifications

.............. $(Sal 5.4 Procedures 5?4-1 5.5 Programs and Mi.nuals 5:541 5.6 Reporting Requirements 5:6i1

[5.7 High Radiation Area................. 5;7#1]

B 2.1 Safety Limits (SLs)

B !! Elf 1 B 2.1.1 Reactor Core SLs................

B!?l!191 B 2.1.2 Reactor Coolant System (RCS) Pressure SL....

B[4231 Bl330?E"GisiFE1Itaquipiiont' '.^.'i ^.'. ".'. '?. '..'. ;. : ;.

8 3.0.1-1 s ~

B 3.0 1 Limit'16g~ Coriditions foe Ope"r'atiori Applicability. '... ~ B' 8.0;1-1

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B 3.0.2 Surveillance Requirement Applicability.........

B(10]!-1 B 3.1 Reactivity Control Systems B$XiF1 B 3.1.1 Shutdown Margin (GDM)

B 31171e1 B 3.1.2 Reactivity Balance...............

B M it-1 BWOG STS iii Rev 1, 04/07/95

SLs 2.1 2

Safety Limits (SLs)QiE(KyQliit@i IEI188 2.1.1 Reactor Core SLs

%~?)~~ temperature shall be s [5080 - (6.5 x 10' pin centerline In MODES 1 and 2, the maximum local fuel MWD /MTU)*F). Operation within this limit is ensured by compliance with the AXIAL POWER IMBALANCE protective limits preserved by the Reactor Protection 1

System setpoints in LCO 3.3.1, " Reactor Protection System (RPS)

Instrumentation," as specified in the COLR.

$E2 In MODES I and 2, the departure from nucleate boiling ratio shall

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be maintained greater than the limits of [1.3 for the BAW-2 correlation and 1.18 for the BWC correlat'ion). Operation within this limit is ensured by compliance with SL 3 and with the AXIAL

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POWER IMBALANCE protective limits preserved by the RPS setpoints in LCO 3.3.1, as specified in the COLR.

SL;3 In MODES I and 2, Reactor Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the SL shown in Figure 2.1.1-1.

l 2.1.2 RCS Pressure SL 5(4 In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained s [2750] psig.

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EI ' BE IIII I

I I

I I

I i ll l

i I

BWOG STS 25-1 Draft Rev 2, 01/25/96

SLs 2.[

2400 This figure for illustration only.

Do not use for operation.

2300 2200

_a g

SAFETY LIMIT MET g

2100 4

SAFETY LIMIT cr n.

m 2000 g

/

O w

1900 0

SAFETY LIMIT VIOLATED 1800

/

1700 I I I I I I I I I I I I I I I I I I 580 590 600 610 620 630 640 REACTOR OUTLET TEMPERATURE ( F)

Figure 2.1.1 1 (page 1 of 1)

Reactor Coolant System Departure from Nucleate Bolling Safety Limits i

BWOG STS 2.@

Draft Rev 2, 01/25/96

$LVs 2.2 SL Violations $(.Ysj With any SL violation, the following actions shall be completed:

I Q[{

Reactor Core SLVs g![{

In MODE 1 or 2, if SL 1 or SL 2 is violated, be in MODE 3 within I hour.

$Vit In MODE 1 or 2, if SL 3 is violated, restore RCS pressure and temperature within limits and be in MODE 3 within I hour.

Q[2 RCS Pressure SLVs SLV'3 In MODE 1 or 2, if SL 4 is not met, restore compliance within

~~

limits and be in MODE 3 within I hour.

SLV ;'4 In MODES 3, 4, and 5, if SL 4 is not met, restore RCS pressure to 5 (2750] psig within 5 minutes.

BWOG STS 2.2-1 Draft Rev 2, 01/25/96

i Reactor Core SLs l

B 2.1.1 B2.[

Safety Limits (SLs)

I B 2.1.1 Reactor Core SLs BASES BACKGROUND GDC 10 (Ref.1) requires that reactor core SLs ensure specified acceptable fuel design limits are not exceeded during steady state operation, normal operational transients, and anticipated operational occurrences (A00s).

This is accomplished by having a departure from nucleate boiling (DNB) design basis, which corresponds to a 95%

probability at a 95% confidence level (95/95 DNB criterion) that DNB will not occur and by requiring that the fuel centerline temperature stays below the melting temperature.

The restrictions of this SL prevent overheating of the fuel and cladding and possible cladding perforation that would result in the release of fission products to the reactor cool ant. Overheating of the fuel is prevented by maintaining the steady state peak linear heat rate (LHR) below the level at which fuel centerline meltira occurs.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime, where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Fuel centerline melting occurs when the local LHR, or power peaking, in a region of the fuel is high enough to cause the fuel centerline temperature to reach the melting point of the fuel.

Expansion of the pellet upon centerline melting may cause the pellet to stress the cladding to the point of failure, allowing an uncontrolled release of activity to the reactor coolant.

Operation above the boundary of the nucleate boiling regime could result in excessive cladding temperature because of the onset of DNB 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.

BWOG STS B2.@-1 Draft Rev 2,01/25/96 m,

n

s RCS Pressure SL B 2.1.2 B 2.[ Safety Limits (SLs)

B 2.1.2 Reactor Coolant System (RCS) Pressure SL BACKGROUND According to 10 CFR 50, Appendix A, GDC 14, " Reactor Coolant Pressure Boundary," and GDC 15, " Reactor Coolant System Design" (Ref.1), the reactor coolant pressure boundary (RCPB) design conditions are not to be exceeded during normal operation nor during anticipated operational occurrences (A00s). GDC 28, " Reactivity Limits" (Ref. 1),

specifies that reactivity accidents including rod ejection do not result in damage to the RCPB greater than limited local yielding.

The design pressure of the RCS is 2500 psig. During normal operation and A00s, the RCS pressure is kept from exceeding the design pressure by mere than 10% in order to remain in accordance with Section III of the ASME Code (Ref. 2).

Hence, the safety limit is 2750 psig. To ensure system integrity, all RCS components are hydrostatically tested at 125% of design pressure prior to initial operation, according to the ASME Code requirements.

Inservice operational hydrotesting at 100% of design pressure is also required whenever the reactor vessel head has been removed or if other pressure boundary joint alterations have occurred.

Following inception of unit operation, RCS components shall be pressure tested, in accordance with the requirements of ASME Code,Section XI (Ref. 3).

APPLICABLE The RCS pressurizer safety valves, operating in conjunction SAFETY ANALYSES with the Reactor Protection System trip settings, ensure that the RCS pressure SL will not be exceeded.

The RCS pressurizer safety valves are sized to prevent system pressure from exceeding the design pressure by more than 10%, in accordance with Section III of the ASME Code for Nuclear Power Plant Components (Ref. 2). The transient that is most influential for establishing the required relief capacity, and hence the valve size requirements and lift settings, is a rod withdrawal from low power.

During the transient, no control actions are assumed except that the safety valves on the secondary plant are assumed to open BWOG STS B2.112.-1 Draft Rev 2,01/25/96

7 _._ ___

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TYPICAL. DOC i

i j

Problems with Brackets:

Is the bracketed material typical information that could be changed for a plant specific TS or is the bracketed material optional, such that the licensee may choose or not choose to use it7 i

This becomes a problem in SGML, as well as WP51 if one wishes to remove

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graphical brackets that are an editing nighmarel i

Solution is to show information a~s typical with L ] s and NOT attempt to j

define it further as optional or as being typical within something that itself is typical!

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SURVEILLANCE REQUIREMENTS 4

SURVEILLANCE FREQUENCY t

j SR 3.7.13.1 Operate each FSPVS train for [2 10 31 days continuous hours with the heaters i

operating or (for systems without j

heaters) 2 15 minutes).

SR 3.7.13.2 Perform required FSPVS filter testing in In accordance accordance with the (Ventilation Filter with the j

Testing Program (VFTP)).

[VFTP) i l

j SR 3.7.13.3 Verify each FSPVS train actuates on an (18] months j

actual or simulated actuation signal.

SR 3.7.13.4 Verify one FSPVS train can maintain a

[18] months on pressure s [ ] inches water gauge with a STAGGERED

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respect to atmospheric pressure during the TEST BASIS l

[ post accident) mode of operation at a flow l

rate s [3000) cfm.

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i i

SR 3.7.13.5 Verify each FSPVS filter bypass damper (18) months can be opened.

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Definitions 1.1 1.1 Definitions 1

LEAKAGE LEAKAGE shall be:

a.

)DENTIfMDLEAKAGE 1.

LEAKAGE, such as that from pump seals or valve packing (except RCP seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; 2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or 3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator (SG) to the Secondary System; I

b.

LMIDENTIF1&D LEAKAGE All LEAKAGE that is not identified LEAKAGE or controlled LEAKAGE; c.

PRESSURE BOUNDARY LEAKAGE l

LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, i

pipe wall, or vessel wall.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning 4

specified in Table 1.1-1 with fuel in the reactor vessel.

NUCLEAR HEAT FLUX HOT WCfEWHEKRLUfHOT?CHANNEQfACTOR$.12)] shall CHANNEL FACTOR M" Et bi the maximum local ' linear powerdenslty in the core divided by the core average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions.

ENGINEERED SAFETY The F.SCRESPONSE! TIME shall be that time interval FEATURE (ESF) RESPONSE from~innEn~tKE~moEit'6 red parameter exceeds its ESF 1

(continued)

BWOG STS 1.1-5 Rev 1, 04/07/95 4

l J

RPS Instrumentation B 3.3.1 B 3.3 INSTRUMENTATION B 3.3.1 Reactor Protection System (RPS) Instrumentation BASES l

BACKGROUND The RPS initiates a reactor trip to protect against i

I violating the core fuel design limits and the Reactor i

Coolant System (RCS) pressure boundary during anticipated i

operational occurrences (A00s). By tripping the reactor, l

the RPS also assists the Engineered Safety Feature (ESF)

Systems in mitigating accidents.

t l

The protection and monitoring systems have been designed to assure safe operation of the reactor. This is achieved by specifying limiting safety system settings (LSSS) in terms of parameters directly monitored by the RPS, as well as the i

LCOs on other reactor system parameters and equipment J

performance.

The LSSS, defined in this Specification as the Allowable Value, in conjunction with the LCOs, establishes the threshold for protective system action to prevent exceeding acceptable limits during Design Basis Accidents (DBAs).

During A00s, which are those events expected to occur one or more times during the unit's life, the acceptable limit is:

li M esie]isti a.

The departure from nucleate boiling ratio (DNBR) shall be maintained above the Safety Limit (SL) value; b.

Fuel centerline melt shall not occur; and c.

The RCS pressure SL of 2750 psia shall not be exceeded.

Maintaining the parameters within the above values ensures that the offsite dose will be within the 10 CFR 20 and 10 CFR 100 criteria during A00s.

iccidents are events that are analyzed even though they are

..ot expected to occur during the unit's life.

The acceptable limit during accidents is that the offsite dose shall be maintained within 10 CFR 100 limits. Meeting the acceptable dose limit for an accident category is considered having acceptable consequences for that event.

(continued) l BWOG STS B 3.3-1 Rev 1, 04/07/95

RPS Instrumentation B 3.3.1 j

BASES l

BACKGROUND RPS Overview B 0PIC (continued)

The RPS consists of four separate redundant protection channels that receive inputs of neutron flux, RCS pressure, 4

RCS flow, RCS temperature, RCS pump status, reactor building (RB) pressure, main feedwater (MFW) pump status, and turbine status.

1 i

Figure (

), FSAR, Chapter [7] (Ref. 1), shows the arrangement of a typical RPS protection channel. A i

protection channel is composed of measurement channels, a manual trip channel, a reactor trip module (RTM), and CONTROL ROD drive (CRD) trip devices.

LCO 3.3.1 provides requirements for the individual measurement channels. These i

channels encompass all equipment and electronics from the l

point at which the measured parameter is sensed through the l

bistable relay contacts in the trip string.

LCO 3.3.2,

" Reactor Protection System (RPS) Manual Reactor Trip,"

LCO 3.3.3, " Reactor Protection System (RPS)-Reactor Trip Module (RTM)," and LCO 3.3.4, " CONTROL ROD Drive (CRD) Trip i

i Devices," discuss the remaining RPS elements.

l The RPS instrumentation measures critical unit parameters j

and compares these to predetermined setpoints.

If the setpoint is exceeded, a channel trip signal is generated.

The generation of any two trip signals in any of the four RPS channels will result in the trip of the reactor.

i The Reactor Trip System (RTS) contains multiple CRD trip l

devices, two AC trip breakers, and two DC trip breaker pairs that provide a path for power to the CRD System.

j Additionally, the power for most of the CRDs passes through electronic trip assembly (ETA) relays. The system has two separate paths (or channels), with each path having either two breakers or a breaker and an ETA relay in series.

Each path provides independent power to the CRDs.

Either path l

can provide sufficient power to operate all CRDs.

Two i

separate power paths to the CRDs ensure that a single failure that opens one path will not cause an unwantea reactor trip.

The RPS consists of four independent protection channels, i

each containing an RTM. The RTM receives signals from its own measurement channels that indicate a protection channel l

trip is required. The RTM transmits this signal to its own two-out-of-four trip logic and to the two-out-of-four logic i

i (continued)

BWOG STS B 3.3-2 Rev 1, 04/07/95 4

l RPS Instrumentation B 3.3.1 i

BASES i

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BACKGROUND RPS Overview (continued)

EMM$CMIMJE0 i

of the RTMs in the other three RPS channels. Whenever any 1

two RPS channels transmit channel trip signals, the RTM logic in each channel actuates to remove 120 VAC power from

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its associated CRD trip breaker.

3 The reactor is tripped by opening circuit breakers that interrupt the power supply to the CRDs. Six breakers are j

installed to increase reliability and allow testing of the trip system. A one-out-of-two taken twice logic is used to j

j interrupt power to the rods.

j The RPS has two bypasses: a shutdown bypass and a channel bypass.

Shutdown bypass allows the withdrawal of safety 4

rods for SDM availability and rapid negative reactivity j

insertion during unit cooldowns or heatups. Channel bypass is used for maintenance and testing. Test circuits in the trip strings allow complete testing of all RPS trip Functions.

The DPS operates from the instrumentation channels discussed i

next. The specific relationship between measurement 4

channels and protection channels differs from parameter to i

parameter. Three basic configurations are used:

a.

Four completely redundant measurements (e.g., reactor coolant flow) with one channel input to each j

protection channel; a

!~

b.

Four channels that provide similar, but not identical, measurements (e.g., power range nuclear instrumentation where each RPS channel monitors a i

different quadrant), with one channel input to each protection channel; and c.

Redundant measurements with combinational trip logic j

outside of the protection channels and the combined j

output provided to each protection channel (e.g., main j

turbine trip instrumentation).

l These arrangements and the relationship of instrumentation 1

channels to trip Functions are discussed next to assist in understanding the overall effect of instrumentation channel failure.

i l

(continued)

BWOG STS B 3.3-3 Rev 1, 04/07/95 4

RPS Instrumentation B 3.3.1 l

BASES

)

BACKGROUND Power Ranae Nuclear Instrumentation

((gTWIC i

(continued)

Power Range Nuclear Instrumentation channels provide inputs to the following trip Functions:

W3331g 1.

EM1f!LMeb1 ]Mi@MM[%gujihjatiggjot l

NuclearOverpowef E!

i Min {@lQA

[ j f [ % ; [ _ C "igh Setpoint; a.

b.

P_f_t?_"_f^"5._5?"LowSetpoint; m

m 7.

Reactor Coolant Pump to Power; 1

Nuclear Overp#if;? f ).E Q f}5s};

ower RCS Flow and Measured AXIAL POWER 8.

IMBALANCE '":

l 9.

Main Turbine Trip (Control Oil Pressure); and 10.

Loss of Main Feedwater (LOMFW) Pumps (Control Oil j

Pressure).

l The power range instrumentation has four linear level channels, one for each core quadrant.

Each channel feeds i

one RPS protection channel.

Each channel originates in a detector assembly containing two uncompensated ion chambers.

l The ion chambers are positioned to represent the top half and bottom half of the core. The individual currents from the chambers are fed to individual linear amplifiers. The sumation of the top and bottom is the total reactor power.

The difference of the top minus the bottom neutron signal is the measured AXIAL POWER IMBALANCE of the reactor core.

Reactor Coolant System Outlet Temoerature j

l l

The Reactor Coolant System Outlet Temperature provides input j

to the following Functions:

2.

RCS High Outlet Temperature; and yijgiuriptistj@

5.

RCS Variable Low Pressure.

l The RCS Outlet Temperature is measured by two resistance elements in each hot leg, for a total of four.

(continued)

BWOG STS B 3.3-4 Rev 1, 04/07/95

RPS Instrumentation B 3.3.1 BASES (continued)

APPLICABLE Each of the analyzed accidents and transients can be SAFETY ANALYSES, detected by one or more RPS Functions. The accident LCO, and analysis contained in (Ref. [ )) takes credit for most RPS l

APPLICABILITY trip Functions. Functions not specifically credited in the accident analysis were qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the l

unit. These Functions are high RB pressure, high temperature, turbine trip, and loss of main feedwater.

l I

These Functions may provide protection for conditions that do not require dynamic transient analysis to demonstrate Function performance. These Functions also serve as backups to Functions that were credited in the safety analysis.

The safety analyses applicable to each RPS Function are discussed next.

TABLELREFERENCE 1.

Nuclear Overnower M % LINED a.

Realeme4ve@iswee-Hioh Setooint i

The Nuclear Overpower-High Setpoint trip provides protection for the design thermal overpower condition based on the measured out of core fast neutron leakage flux.

The Nuclear Overpower-High Setpoint trip initiates a reactor trip when the neutron power reaches a predefined setpoint at the design overpower limit.

Because THERMAL POWER lags the neutron power, tripping when the neutron power reaches the design overpower will limit THERMAL POWER to a maximum value of the design overpower.

Thus, the Nuclear Overpower-High Setpoin't. trip protects against violation of the DNBR and fuel centerline melt SLs.

b.

M Low Setooint While in shutdown bypass, with the Shutdown Bypass RCS High Pressure trip OPERABLE, the Nuclear Overpower-Low Setpoint trip must be reduced to 5 5% RTP. The low power setpoint, in i

conjunction with the lower Shutdown Bypass RCS l

High Pressure setpoint, ensure that the unit is (continued)

)

i BWOG STS B 3.3-5 Rev 1, 04/07/95

- = - -.- -.-

RPS Instrumentation i

B 3.3.1 1

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BASES (continued) i ACTIONS B.13EiF6r3 For Required Action B.1 and Required Action B.2, if one or more Functions in two protection channels become inoperable, one of two inoperable protection channels must be placed in trip and the other in bypass. These Required Actions place all RP3 Functions in a one-out-of-two logic configuration and prevent bypass of a second channel.

In this i

configuration, the RPS can still perform its safety functions in the presence of a random failure of any single i

channel. The I hour Completion Time is sufficient time to i

perform Required Action B.1 and Required Action B.2.

l 1

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$5e7A71[a%fs?

L1 i

i Required Action C.1 directs entry into the appropriate Condition referenced in Table 3.3.1-1.

The applicable l

Condition referenced in the table is Function dependent.

j SURVEILLANCE The SRs for each RPS Function are identified by the SRs REQUIREMENTS column of Table 3.3.1-1 for that Function.

SR 3.3.1.1" " **" *~ '*

j Performance of the CHANNEL CHECK once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ensures i

that a gross failure of instrumentation has not occurred. A j

CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other j

channels.

It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value.

C 111f1!2 S R M R313(II11ab6 M SR 3.3.1.3 A comparison of power range nuclear instrumentation channels against incore detectors shall be performed at a 31 day (continued)

BWOG STS B 3.3-6 Rev 1, 04/07/95

i j

RPS Instrumentation 3.3.1 j

Table 3.3.1 1 (page 1 of 1) 4 Reactor Protection System Instruentation h

APPLICABLE Com ITIONS a

i IRBES OR REFERENCED OTHER FRtlBl 3

SPECIFIED REGUIRED SURVEILLANCE ALLOWABLE I

FUNCTION COW ITIONS ACTION C.1 REGUIRE8ENTS VALUE 1.

NucteerOverpower3 e.

Righ setpoint 1,2(*)

D SR 3.3.1.1 s [104.911 RTP BR 3.3.1.2 1

SR 3.3.1.5 4

SR 3.3.1.7 b.

tow setpoint 2(b) 3(b)

E st 3.3.1.1 s 55 RTP k

at 3.3.1.5

'j

40) 5, D3 sa 3.3.1.7

)

2.

RCs Wish outlet Temperature 1,2 0

st 3.3.1.1 s (618)*F SR 3.3.1.4 SR 3.3.1.6 3.

RCS Nigh Pressure 1,2 D

st 3.3.1.1 5 (2355) psig i

st 3.3.1.4 i

SR 3.3.1.6 SR 3.3.1.7 4.

RCs Low Pressure 1,2(a)

D st 3.3.1.1 m (18003 pois 54 3.3.1.4 SR 3.3.1.6 st 3.3.1.7

5. RCs Verlebte Low Pressure 1,2(a)

D st 3.3.1.1 t ([11.59) eT t*

SR 3.3.1.4 (5037.8))psYe SR 3.3.1.6 6.

Reactor Building High 1,2,3(*)

D st 3.3.1.1 5 [4] pois Pressure SR 3.3.1.4 st 3.3.1.6 7.

Reector Coolant Pop to 1,2(a)

D

$R 3.3.1.1 (51% RTP with 5 2 Power 3R 3.3.1.4 pe ps operating 3R 3.3.1.6 st 3.3.1.7 8.

Ilucteer Overpower RCS Flow 1,2(a)

D st 3.3.1.1 Nuclear Overpower RCs and Measured AXIAL POWER st 3.3.1.3 Flow and AX!AL POWER INEALANCE st 3.3.1.5 INsALANCE setpoint SR 3.3.1.6 envelope in COLR st 3.3.1.7 9.

Noin Turbine Trip (Control t (4511 RTP F

SR 3.3.1.1 t (451 psig i

Dit Pressure) st 3.3.1.4 st 3.3.1.6

10. Loss of Nein Feeduster t [1511 RTP G

st 3.3.1.1 t (55) psis Po ps (Controt oil SR 3.3.1.4 Pressure)

SR 3.3.1.6

11. shutdown Bypese RCs Nigh 2(b) 3, IDI E

st 3.3.1.1 s [17202 psis Pressure SR 3.3.1.4 4(b) 5(b) st 3.3.1.6 (e) Isen not in shutdoom bypass operation.

(b) During shutdown bypass operation with any CRD trip breekers in the closed position and the CR0 System capable of rod withdrawet.

(c) With any CRD trip breaker in the closed position and the CRD System capable of rod withdrawal.

BWOG STS 3.3-5 Rev 1, 04/07/95

Site Location 4.1 i

4 Design Features 4.1 Site Location 1

i

[ Text description of site location.]

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l BWOG STS 4.1-Rev1,10/28/96 j

.. ~.. -

Reactor Core 4.2 4.2 Reactor Core 4.2.1 Fuel Attannhlies The reactor shall contain [177] fuel assemblies. Each assembly j

shall consist of a matrix of [Zircalloy or ZIRLO] fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U0 ) as fuel material. Limited substitutions of zirconium 2

4 alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be 4

i used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel i

safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in non11miting core regions.

4.2.2 control Rods The reactor core shall contain [60] safety and regulating and [8]

axial power shaping CONTROL RODS. The control material shall be i

[ silver indium cadmium, boron carbide, or hafnium metal] as approved by the NRC.

6 4

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BWOG STS 4.2-Rev1,10/28/96

Fuel Storage 4.3 4.3 Fuel Storage 4.3.1 critica11tv 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

4

a. Fuel assemblies having a maximum U-235 enrichment of

[4.5] weight percent;

b. k,,, s 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described 4

j in [Section 9.1 of the FSAR];

c. A nominal [

] inch center to center distance between fuel assemblies placed in [the high density fuel storage racks];

d. A nominal [

] inch center to center distance between fuel assemblies placed in [the low density fuel storage racks];

i

e. New or partially spent fuel assemblies with a discharge burnup in the " acceptable range" of 3.7.17-1 may be allowed unrestricted storage in [either] fuel storagerack(s);and
f. New or partially spent fuel assemblies with a l

discharge burnup in the " unacceptable range" of 3.7.17-1 will be stored in compliance with the NRC approved [ specific document containing the analytical methods, title, date, or specific configuration or figure].

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of

[4.5] weight percent;

b. k,,,5 0.95 is fully flooded with unborated water, which includes an allowance for uncertainties as described in [Section 9.1 of the FSAR];

BWOG STS 4.3-Rev1,10/28/96

Fuel Storage 4.3 J

c. k,n 5 0.98 if moderated by aqueous foam, which includes l

an allowance for uncertainties as described in l

[Section 9.1 of the FSAR]; and

d. A nominal [21.125] inch center to center distance l

between fuel assemblies placed in the storage racks.

i 4.3.2 orainage 5

l The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation [138 ft 4 i

inches].

i 4.3.3 canacity The spent fuel storage pool is designed and shall be maintained 1

with a storage capacity limited to no more than [1357] fuel assemblies [and six failed fuel containers].

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BWOG STS 4.3-Rev1,10/28/96

_. _. _. _ ~.. _...

Improved standard T:chnical specifice.tions http://www.nre. gov /NRR/sts/sts.htm l

Improved Standard Technical Specifications l

l NUREG Documents Revision 1 of the Standard Technical Specifications are published in five NUREG documents as follows:

4 i

TITLE

  • "-DOCUMENT-i Standard Technical Specifications Babcock and Wilcox Plants ""*" NUREG-1430 l

Standard Technical Specifications Waeinghause Plants """""" NUREG-1431 l

Standard Technical Speci6 cations Combustion F=p=:-:ig Plants "* NUREG-1432 j

Standard Technical Specifications General Electric Plants, BWR/4 *" NUREG-1433 j

Standard Technical Specifications General Electric Plants, BWR/6 "* NUREG-1434 i

l Each NUREG consists of 3 Volumes, issued as Rev.1, APRIL 1995. Filenames ending in " ZIP" are i

compressed Wordperfect 5.1 files. Download (Load to Local Disk) PKZ204G.EXE from PKWare. Inc.

to uncompress zipped files.

Drafts of Revision 2 of the Standard Technical Specifications are included at the top of the list for each of the above NUREG file lists.

A set of modified Bases documents is included at the bottom of the lists for each of the vendor NUREGs. The modified Bases documents are in a format that permits pagination of the document. This version of the Bases documents is listed as 1A since the text of these documents is the same as Revision 1 of the NUREG documents.

(Some files have been updated to reflect approved Draft Rev 2 changes.) The revised bases were created using Wordperfect Macros which may have application for plant TS Bases. The macros and a writeup on their use is as follows:

-Description of File-FILENAME---

Macros used to create Rev 1 A Bases DocumetsMACRO. ZIP ( 60 kbytes)

Writeup on Macros used to create Rev 1 A DocumentsWRITEUP. ZIP ( 78 kbytes)

Plants Converting to Standard Technical Specifications Schedule for Conversions (Table Format)

Schedule for Conversions (non-Table Format)

Background / Generic Information 1 of 7 10/17/96 10:20:46

m._._._

j Improved standard Technical specifications ht tp : / /ww.nre. g ov/NRR/sts /sts.htm i

STS individual Ses: Fdename protocol j

Compressed (zipped) Ses: Fdename protocol namment infonnation + printer test README2. ZIP (63kbytes) l Generic Informaation l

l

-Description ofFdes and F"- ----

Appendet J, Tech Specs Option B, Interim Model STS APPJOPTB.ZF ( 68 kbytes)

SGML Tech Spec Inforniation l

-Description of Files and Filenames-i l

Dam===r Type Defmition(DTD) for SGML Tech Specs (Modl5) TSMOD15.DTD ( 30 kbytes) i Cais Table Entity for SGML Tech Specs CALS-TBL.ENT ( 8 kbytes) l Character Set Entity for SGML Tech Specs CHAR-SET.ENT ( 2 kbytes) j Front Matter Entity for SGML Tech Specs FRONT.ENT ( 2 kbytes) i Reat Matter Endty for SGML Tech Specs REAR.ENT ( l kbytes) i General Electric Plants, BWR/4, NUREG-1433

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Cover Sheets for BWR/4 STS 4S COVER HP (149 kbytes)

Table ofContents 4ST TCRI.ZT ( 12 kbytes)

CHAPTER 1.0 Spec: Use and A alication 4ST10 SRI. ZIP ( 21 kbytes)

CHAPTER 2.0 Spec: Safety Limits 4ST20SR1.ZF ( 2 kbytes)

CHAPTER 2.0 Bases: Safety Limits 4ST2IBRI. ZIP ( 10 kbytes)

CHAPTER 3.0 LCO: Appliability 4ST30LRI.ZF ( 4 kbytes)

CHAPTER 3.0 Bases: Appliability 4ST30BR1. ZIP ( lI kbytes)

CHAPTER 3.1 LCO: Reactivity Control Systems 4ST31LR1. ZIP (102 kbytes)

CHAPTER 3.1 Bases: Reactivity Control Systems 4ST31BRI. ZIP ( 45 kbytes)

CHAPTER 3.2 LCO: Power Distribution Limits 4ST32LRI. ZIP ( 9 kbytes)

CHAPTER 3.2 Bases: Power Distribution Limits 4ST32BR1. ZIP ( 19 kbytes)

CHAPTER 3.3 LCO: Instrumentation 4ST33LR1.ZF ( 65 kbytes)

CHAPTER 3.3 Bases: Ins *rumentation 4ST33BRI. ZIP (165 kbytes)

CHAPTER 3.4 LCO: Reactor Coolsnt System 4ST34LR1.ZF ( 30 kbytes)

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CHAPTER 3.5 LCO: Emergency Core CM S

CHAPTER 3.5 Bases: Emergency Core Csc!ing Systems 4ST35BR1. ZIP ( 23 kbytes)

CHAPTER 3.6 LCO: Containment Systems 4ST36LRI. ZIP ( 48 kbytes)

CHAPTER 3.6 Bases: Containment Systems 4ST36BR1. ZIP (106 kbytes)

CHAPTER 3.7 LCO: Plant Systems 4ST37LRI. ZIP ( l8 kbytes)

CHAPTER 3.7 Bases Plant Systems 4ST37BRI. ZIP ( 35 kbytes) 2 of 7 10/17/96 10:21:02-

. =

f Improvsd Standard Technical Specificatiens http://www.nre.Dov/NRR/sts/sts.htm j

i CHAPTER 3.8 LCO: Electrical Power Systems 4ST38LR1. ZIP ( 31 kbytes) i CHAPTER 3.8 Bases: Electrical Power Systems 4ST38BR1. ZIP ( 73 kbytes) l CHAPTER 3.9 LCO: Refueling Operations 4ST39LRI. ZIP ( 17 kbytes) l CHAPTER 3.9 Bases: PMaa Operations (ST39BRI. ZIP ( 33 kbytes)

CHAPTER 3.10 LCO: Special Operations 4ST3ALR1. ZIP ( 24 kbytes) j CHAPTER 3.10 Bases: Special Operations 4ST3 ABRI. ZIP ( 43 kbytes)

CHAPTER 4.0 Spec: Design Features 4ST40 SRI. ZIP ( 3 kbytes)

+

CHAP 1ER 5.0 Spec: Administrative Controls 4ST50 SRI. ZIP ( l8 kbytes)

Revision 1A Bases Documents:

1 CHAPTER 2.0 Bases: Safety Limits 4S21BRI A. ZIP ( 11 kbytes)

CHAPTER 3.0 Bases: ApA4h 4S30BR1 A. ZIP ( lI kbytes) l CHAPTER 3.1 Bases: Reactivity Control Systems 4S31BR1 A. ZIP ( 49 kbytes)

}

CHAPTER 3.2 Bases: Power Distribution Limits 4S32BR1 A. ZIP ( 21 kbytes) l CHAPTER 3.3 Bases Instrumentation 4S33BRI A. ZIP (166 kbytes)

CHAPTER 3.4 Bases: Reactor Coolant System 4S34BRI A. ZIP ( 59 kbytes) 3 CHAPTER 3.5 Bases: Emergency Core Cooling Systems 4S35BRI A. ZIP ( 24 kbytes)

CHAPTER 3.6 Bases: Containment Systems 4S36BR1 A. ZIP (109 kbytes)

CHAPTER 3.7 Bases: Plant Systems 4S37BR1 A ZIP ( 38 kbytes)

CHAPTER 3.8 Bases: Electrical Power Systems 4S38BR1 A ZIP ( 74 kbytes)

CHAPTER 3.9 Bases: Refueling Operations 4S39BR1 A7IP ( 37 kbytes) j CHAPTER 3.10 Bases: Special Operations 4S3 ABR1 A. ZIP ( 48 kbytes) i i

i General Electric Plants, BWR/6, NUREG-1434

-Description ofFile FILENAME--

j First set ofDraR Rev 2 files 6SD2-1. ZIP ( 10 kbytes) l First set ofDraR Rev 2 editorial change files 6SD2-1E. ZIP ( 14 kbytes) i j

Cover Sheets for BWR/6 STS 6S COVER. ZIP (149 kbytes) l Table of Contents 6ST TCRI. ZIP ( 12 kbytes) 1 CHAPTER 1.0 Spec: Use and Application 6ST10 SRI. ZIP ( 21 kbytes)

CHAPTER 2.0 Spec: Safety Limits 6ST20 SRI. ZIP ( 2 kbytes) 3 CHAPTER 2.0 Bases: Safety Limits 6ST21BRI. ZIP ( 10 kbytes)

CHAPTER 3.0 LCO: Appliability 6ST30LRI. ZIP ( 5 kbytes) i CHAPTER 3.0 Bases: Appliability 6ST30BR1. ZIP ( 11 kbytes)

I CHAPTER 3.1 LCO: Reactivity Control Systems 6ST31LR1. ZIP ( 82 kbytes) j CHAPTER 3.1 Bases: Reactivity Control Systems 6ST31BR1. ZIP ( 45 kbytes) j CHAPTER 3.2 LCO: Power Distribution Limits 6ST32LRI. ZIP ( 7 kbytes) i CHAPTER 3.2 Bases: Power Distribution Limits 6ST32BRI. ZIP ( l8 kbytes)

CHAPTER 3.3 LCO: Instrumentation 6ST33LR1. ZIP ( 69 kbytes)

CHAPTER 3.3 Bases: Instrumentation 6ST33BR1. ZIP (173 kbytes)

CHAPTER 3.4 LCO: Reactor Coolant System 6ST34LR1. ZIP ( 27 kbytes)

CHAPTER 3.4 Bases: Reactor Coolant System 6ST34BR1. ZIP ( 58 kbytes)

CHAPTER 3.5 LCO: Emergency Core Cooling Systems 6ST35LR1. ZIP ( 10 kbytes)

CHAPTER 3.5 Bases: Emergency Core Cooling Systems 6ST35BR1. ZIP ( 21 kbytes) 3 of 7 10/17/96 10:21:08

Improved standard T:chnical Specifications http://www.nre. gov /NRR/sts/sts.htm i

CHAPTER 3.6 LCO: Containment Systems 6ST36LRI. ZIP ( 61 kbytes) l CHAPTER 3.6 Bases: Containment Systems 6ST36BR1. ZIP (128 kbytes)

CHAPTER 3.7 LCO: Plant Systems 63T37LR1. ZIP ( 16 kbytes) j CHAPTER 3.7 Bases: Plant Systems 6ST37BR1. ZIP ( 32 kbytes) l CHAPTER 3.8 I.CO: Electrical Power Systems 6ST38LR1. ZIP ( 32 kbytes)

CHAPTER 3.8 Bases: Electrical Power Systems 6ST38BR1. ZIP ( 73 kbytes)

CHAPTER 3.9 LCO: Rddia Operations 6ST39LRI.ZF ( 17 kbytes)

CHAPTER 3.9 Bases: RdWkg Operations 6ST39BRI.ZF ( 31 kbytes)

CHAPTER 3.10 LCO: Special Operations 6ST3 ALRI.ZF ( 24 kbytes)

CHAPTER 3.10 Bases: Special Operations 6ST3 ABRI. ZIP ( 43 kbytes)

CHAPTER 4.0 Spec: Design Features 6ST40SR1. ZIP ( 3 kbytes)

CHAPTER 5.0 Spec: Administrative Controls 6ST50SR1.ZF ( l8 kbytes)

Revision 1A Bases Documents:

CHAPTER 2.0 Bases: Safety Limits 6S21BR1 A ZIP ( 11 kbytes)

CHAPTER 3.0 Bases: Appliability 6S30BR1 A. ZIP ( 12 kbytes)

CHAPTER 3.1 Bases: Reactivity Control Systems 6S31BRI A. ZIP ( 49 kbytes)

CHAPTER 3.2 Bases: Power Distribution Limits 6S32BR1 A. ZIP ( 21 kbytes)

CHAPTER 3.3 Bases: Instrumentation 6S33BR1 A. ZIP (174 kbytes)

CHAPTER 3.4 Bases: Reactor Coolant System 6S34BRI A. ZIP ( 64 kbytes)

CHAPTER 3.5 Bases: Emergency Core Cooling Systems 6S35BRI A ZIP ( 23 kbytes)

CHAPTER 3.6 Bases: Containment Systems 6S36BR1 A ZIP (138 kbytes)

CHAPTER 3.7 Bases: Plant Systems 6S37BRI A. ZIP ( 33 kbytes)

CHAPTER 3.8 Bases: Electrical Power Systems 6S38BRI A. ZIP ( 74 kbytes)

CHAPTER 3.9 Bases: Refueling Operations 6S39BRI A.ZF ( 35 kbytes)

CHAPTER 3.10 Bases: Special Operations 6S3 ABR1 A. ZIP ( 47 kbytes) i Babcock and Wilcox Plants, NUREG-1430

-Description of File----

FILENAME--

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Cover Sheets for BWOG STS BS_ COVER. ZIP (147 kbytes)

TABLE OF CONTENTS BST_TCR1 N ( lI kbytes)

CHAPTER 1.0 Spec: Use and Application BST10SR1.ZF ( 20 kbytes)

CHAPTER 2.0 Spec: Safety Limits BST20SR1. ZIP ( 21 kbytes)

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CHAPTER 3.0 I.CO: Appliability BST30LR1. ZIP ( 4 kbytes)

CHAPTER 3.0 Bases: Appliability BST30BRI. ZIP ( lI kbytes)

CHAPTER 3.1 LCO: Reactivity Control Systems BST31LRI. ZIP ( 20 kbytes)

CHAPTER 3.1 Bases Reactivity Control Systems BST31BRI. ZIP ($1 kbytes)

CHAPTER 3.2 LCO: Power Distribution Ilmits BST32LRI. ZIP ( 11 kbytes)

CHAPTER 3.2 Bases: Power Distribution Limits BST32BR1. ZIP ( 79 kbytes)

CHAPTER 3.3 LCO: Instrumentation BST33LR1. ZIP ( 49 kbytes)

CHAPTER 3.3 Bases: Instrumentation BST33BR1. ZIP (127 kbytes)

CHAPTER 3.4 LCO: Reactor Coolant System BST34LRI.ZF ( 58 kbytes) 4 of 7 10/17/96 10:21:17

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CHAPTER 3.5 LCO: Emergency Core Cooling Systems BST35LRI. ZIP ( 10 kbytes)

CHAPTER 3.5 Bases: Emergency Core Cooling Systems BST35BRI. ZIP ( 25 kbytes)

CHAPTER 3.6 LCO: Containment Systems BST36LR1. ZIP ( 19 kbytes)

CHAPTER 3.6 Bases: Containment Systems BST36BRI. ZIP ( 47 kbytes)

CHAPTER 3.7 LCO: Plant Systems BST37LRI. ZIP (1-9 kbytes)

CHAPTER 3.7 Bases: Plant Systems RST37BRI. ZIP ( 81 kbytes)

CHAPTER 3.8 LCO: Electrical Power Systems BST38LRI. ZIP ( 30 kbytes)

CHAPTER 3.8 Bases Electrical Power Systems BST38BRI. ZIP ( 72 kbytes)

CHAPTER 3.9 LCO: R=A=% Operations BST39LRI.ZF ( 12 kbytes)

CHAPTER 3.9 Bases: Refueling Operations BST39BRI. ZIP ( 24 kbytes)

CHAPTER 4.0 Spec: Design Features BST40 SRI. ZIP ( 3 kbytes)

CHAPTER 5.0 Spec: Administrative Controls BST50SR1. ZIP ( 19 kbytes)

Revision IA Bases Documents:

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CHAPTER 3.0 Bases: Appliability BS]DBR1 A. ZIP ( lI kbytes)

CHAPTER 3.1 Bases: Reactivity Control Systems BS31BRI A. ZIP ( 56 kbytes)

CHAPTER 3.2 Bases: Power Distribution Limits BS32BRI A. ZIP ( 82 kbytes)

CHAPTER 3.3 Bases: Instrumentation BS33BR1 A ZIP (132 kbytes)

CHAPTER 3.4 Bases: Reactor Coolant System BS34BRI A ZIP ( 90 kbytes)

CHAPTER 3.5 Bases: Emergency Core Cooling Systems BS35BRI A. ZIP ( 27 kbytes)

CHAPTER 3.6 Bases: Containment Systems BS36BRI A ZIP ( 49 kbytes)

CHAPTER 3.7 Bases: Plant Systems BS37BRI A. ZIP ( 81 kbytes)

CHAPTER 3.8 Bases: Electrical Power Systems BS38BRI A. ZIP ( 73 kbytes) l CHAPTER 3.9 Bases: Refueling Operations BS39BR1 A ZIP ( 27 kbytes)

Combustion Engineering Plants, NUREG-1432

-Description of File-

F1ENAME--

First set of Draft Rev 2 files CSD2-1. ZIP ( 66 kbytes)

Cover Sheets for CEOG STS CS_ COVER 75 (148 kbytes)

Table of Contents CST TCRI 79 ( 14 kbytes) l CHAPTER 1.0 Spec: Use and Application CST 10 SRI. ZIP ( 20 kbytes)

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CHAPTER 3.1 LCO: Reactivity Control Systems CST 31LRI. ZIP ( 57 kbytes)

CHAPTER 3.1 Bases: Reactivity Control Systems CST 31BRI. ZIP (110 kbytes)

CHAPTER 3.2 LCO: Power Distribution Limits CST 32LRI. ZIP ( 24 kbytes) l CHAPTER 3.2 Bases: Power Distribution Limits CST 32BRI. ZIP ( 64 kbytes)

CHAPTER 3.3 LCO: Instrumentation CST 33LRI. ZIP (145 kbytes) i CHAPTER 3.3 Bases: Instmmentation CST 33BRI. ZIP (313 kbytes)

(

CHAPTER 3.4 LCO: Reactor Coolant System CST 34LRI. ZIP ( 62 kbytes) 5 of 7 10/17/96 10:21:25

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Improv standard Technical Specifications http://www.nrc. gov /NRR/sts/sts.htm i

CHAP 1ER 3.4 Bases: Reactor Coolant System CST 34BRI. ZIP ( 86 kbytes) l CHAPTER 3.5 LCO: Emergency Core Cooling Systems CST 35LRI. ZIP ( l I kbytes) i CHAPTER 3.5 Bases Emergency Core Cooling Systems CST 35BRI. ZIP ( 30 kbytes) l CHAPTER 3.6 ILO: Containment Systems CST 36LRI. ZIP ( 36 kbytes) i CHAP 1ER 3.6 Bases: Ca*Mamad Systems CST 36BRI. ZIP ( 88 kbytes) l CHAPTER 3.7 LCO: Plant Systems CST 37LRI. ZIP ( 63 kbytes) l CHAPTER 3.7 Bases: Plant Systems CST 37BRI. ZIP ( 87 kbytes) j CHAPTER 3.8 LCO: Electrical Power Systems CST 38LR1. ZIP ( 30 kbytes)

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CHAPTER 4.0 Spec: Design Features CST 40 SRI. ZIP ( 3 kbytes)

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j i

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{

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{

CHAPTER 3.7 Bases: Plant Systems CS37BR1 A. ZIP ( 94 kbytes) l CHAPTER 3.8 Bases: Electrical Power Systems CS38BR1 A. ZIP ( 72 kbytes) l CHAPTER 3.9 Bases: Refueling Operations CS39BR1 A. ZIP ( 30 kbytes) i Westinghouse Plants, NUREG-1431 l

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First set of Draft Rev 2 editorial change files WSD2-1E. ZIP ( 40 kbytes)

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CHAPTER 3.0 LCO: Appliability WST30LR1. ZIP ( 4 kbytes) i CHAPTER 3.0 Bases: Appliabdity WST30BR1. ZIP ( 10 kbytes)

CHAPTER 3.1 LCO: Reactivity Control Systems WST31LRI. ZIP ( 40 kbytes) l CHAPTER 3.1 Bases: Reactivity Control Systems WST31BRI. ZIP ( 87 kbytes)

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6 of 7 10/17/96 10:21:33

Improved standard Tech'nic31 Specifications http://www.nrc. gov /NRR/sts/sts.htm

  • j e i

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CHAPTER 3.5 LCO: Emergency Core Cooling Systems WST35LRI. ZIP ( 14 kbytes)

CHAPTER 3.5 Bases. Emergency Core Cooling Systems WST35BRI. ZIP ( 35 kbytes)

]

CHAPTER 3.6 ILO: Containment Systems WST36LRI. ZIP ( 88 kbytes)

CHAPTER 3.6 Bases: Coa *Amant Systems WST36ER12[E (162 kbytes)

CHAPTER 3.7 LCO: Plant Systems WST37LRI. ZIP ( 56 kbytes)

CHAPTER 3.7 Bases: Plant Systems WST37BRI. ZIP ( 87 kbytes)

CHAPTER 3.8 LCO: Electrical Power Systems WST38LRI. ZIP ( 30 kbytes)

CHAPTER 3.8 Bsses: Electrical Power Systems WST38BRI. ZIP ( 73 kbytes)

CHAPTER 3.9 LCO: Refueling Operations WST39LRI. ZIP ( 13 kbytes)

CHAPTER 3.9 Bases: Refueling Operations WST39BRI. ZIP ( 26 kbytes)

CHAPTER 4.0 Spec: Design Features WST40SR1. ZIP ( 3 kbytes)

CHAPTER 5.0 Spec: Administrative Controls WST50 SRI. ZIP ( 18 kbytes)

Revision IA Bases Documents:

CHAPTER 2.0 Bases: Safety Limits WS21BR1 A ZIP ( 70 kbytes)

CHAPTER 3.0 Baces: Appliability WS30BR1 A. ZIP ( lI kbytes)

CHAPTER 3.1 Bases: Reactivity Control Systems WS31BR1 A ZIP (100 kbytes)

CHAPTER 3.2 Bases: Power Distribution Limits WS32BR1 A. ZIP (142 kbytes)

CHAPTER 3.3 Bases: Instrumentation WS33BR1 A. ZIP (148 kbytes)

CHAPTER 3.4 Bases: Reactor Coolant System WS34BR1 A. ZIP (108 kbytes)

CHAPTER 3.5 Bases: Emergency Core Cooling Systems WS35BRI A. ZIP ( 38 kbytes)

CHAPTER 3.6 Bases: Containment Systems WS36BR1 A. ZIP (175 kbytes)

CHAPTER 3.7 Bases: Plant Systems WS37BR1 A. ZIP ( 81 kbytes)

CHAPTER 3.8 Bases: Electrical Power Systems WS38D11 A ZIP ( 75 kbytes)

CHAPTER 3.9 Bases: Refueling Operations WS39BR1 A ZIP ( 30 kbytes)

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4 DIFFERENCES BETWEEN WP51 Als SGML F3RNATTED STS l

l 1.

Chapters are numbered with as single digit, 1 2 3 4 5 and the two 3.0

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Sections are replaced by two Specifications 3.0.1 and 3.0.2, with same content as before, under a single Section 3.0 named General

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Requirements.

(Title could be changed)

REASON: Two sections with same number are incompatible with hyper links j

based on content.

2.

Defined Terms are not listed in the term column with their associated acronym. Hence the text of definitions do not start using the acronym 4

i form, but rather the defined ters, that includes acronyms as they previously appeared in the term column. The acronym form of the defined term is tagged as an alternate ters, < alt. tem >, but style sheets would l

not print the alternate terms under the term column, but the acronym forms are included as part of the text for the definition as noted j

previously. Style sheets for editing the sgal instances would show the alternate terms so that the could be readily edited.

i REASON: Hyper links are based on term content, which must be the same l

as where terms are used. Alternate terms provide the content reference for hyper links for the acronym use.

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3.

Plural forms of defined terms are NOT use by adding a lower case s, e.g.

APSR (Axial Power Shaping Rod) are APSRS not APSRs. This applies as l

well as to the spelled out version in words, e.g., not: AXIAL POWER l

SHAPING RODS. Where a define term in singular, an alternate term exists for the plural versions and as well as one for the plural acronym form.

i REASON Maintain consistent content of term content for hyper links, j

that include plural versions.

i 4.

LEAKAGE definitions for Identified, Unidentified, and Pressure Boundary, j

were converted to all capital letter form, e.g.,

IDENTIFIED LEAKAGE.

1 REASON Maintain consistent content of term content for hyper links.

5.

Mdash dual term for OPERABILITY-0PERABLE was replace with OPERABILITY under the defined term column, and "(or have OPERABILITY)" was added as part of the definition. OPERABLE is an alternate tem.

(OPERABILITY in the Term column will be replace with OPERABLE so tigat the term matches the first usage in the definition of it.)

i' 6.

Currently (B&W STS), safety limits are numbered with 4 digit numbers j

that follow a numbered spec title or carry the spec title number in the case of the RCS Pressure SL. The text addresses them as "SL x.x.x.x" use the SL acronym. The actual numbers are 2.1.1.1, 2.1.1.2, 2.1.1.3, and 2.1.2.

For simplicity, these were renumber as 1, 2, 3, and 4 and

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when listed, include the "SL" as they are referenced in the text.

j REASON: For consistency with hyper linking, numbers that are referenced i

such as LCO 3.0.1 and SR 3.6.5.2, are included in the instances as just 4

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i numbers where the style sheets know that these number are to be output with their prefix, LCO or SR. Hence, tags such as <1co.num.ref>,

<sr.num.ref>, and <s1.num.ref> are used, that latter which likewise outputs the "SL" preceding the number where referenced in the text. (It is preferable, but not absolutely necessary, that the specification of the safety limit includes SL prefix to its number.

Given this situation, there would be no reason for the initial specification of SL to be displayed as mulit digit numbers, a separate one of which does not l

exits for the RCS Pressure SL and would be necessary. E.g.:

f 2.1.2 RCS Pressure Limit

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SL 2.1.2.1 In MODES 1, (etc. with the existing spec).

Four digits could be used, but it makes it much harder to follow. For example, action numbers within 1cos could have carried the spec number j

so that you would have had actions like A.3.1.1.1 instead of A.1 for LCO 3.1.1, and all action numbers would have been unique, but it would have used up a lot of space within action table for very little gain, and l

would be harder to follow.

7.

Currently the safety limit violations are just spec numbers that in the i

Bases are put in an heading such as the following:

1 "The following SL violation responses are applicable to the reactor core SL:

2.2.1 and 2.2.2 (Discussion)

The only place where these numbers appear is in the initial spec entries and in the bases as the header line shown above. One could continue this, however, for consistency with SL, the term SLVs was coined to reference these items.

Likewise they were numbered consecutively, I thru n.

If there is a better choice rather than "SLV" one could certainly be used. However, some identification is desirable for these items that editors can relate to when choosing the tag for entering the data in the authoring environment. Hence it is desirable to have something that will relate to the data, e.g., SLV 1 is tagged

<SLV.NUM>1</SLV.NUM> where it is specified, and

<SLV.NUM.REF>1</SLV NUM.REF> where it is referenced. Multi-digit numbers, while undesirable, could be used and are got an impediment to hyper linking, nor would the acronym prefix be an' absolute requirement.

The major difference with these numbers, as compared to SLs, is that they do not appear in text usage.

I 8.

Safety limits are subdivided into sections for Reactor Core and RCS Pressure. This subdivision within the SL Section was carried over to the Section on SL Violations.

REASON: This was done for clarity and consistency of subsections and is necessary for consistent SGML tagging of these items and allows auto generation of the index. For example, the current index for safety

i limits is:

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2.0 SAFETY LIMITS (SLs) 2.0-1 2.1 SLs 2.0-1 1

2.2 SL Violations 2.0-1 This would be expanded in the electronic versions, and perhaps iq the paper version, of SGML STS to be:

2 Safety Limits (SLs) and SL Violations 2.1-1 2.1 SLs 2.1-1 2.1.1 Reactor Core SLs.................... 2.1-1 1

2.1.2 RCS Pressure SL 2.1-1 2.2 SL Violations (SLVs)..................

2.2-1 2.2.1 Reactor Core SLVs 2.2-1

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2.2.2 RCS Pressure SLVs 2.2-1 Thus, if mulit-digits were retained for SLVs, they would change from 3 to 4 digit numbers, which would be the same if retained for SLs.

9.

Capitalization:

In titles, notably Chapter as show above, and for i

defined terms in titles, lower case is used for second and subsequent l

letters in a word.

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REASON: Space saving for electronic indexing. This is not an SGML l

requirement, hence, is something that could be changed for TS j

j conversions. For present SGML demonstration, this style is used.

1 10.

Horizontal rules between various subsections and at the top and bottom of every page are not included in SGML style sheets, and would not be included in SGML instances as a special tag (calling out ruled lines) which if used would also need special rules on usage within style sheets. Note: The line rule practice was not incorporated in Sections 3.1 thru 3.n of the existing STS.

j REASON: Unduly complicates SGML implementation.

11.

Action Tables and Surveillance Requirements Tables do not use an extra i

row just to include continuation notation at the bottom of a table where it continues on the next page.

REASON: Unduly complicates SGML implementation.,.,

j 12.

File management: Files are by Section for all Chapters except 3 where they are by specification. Currently, STS Chapters 2, 4, and 5 are l

single files.

j REASON: Chapter files unduly complicates SGML implementation.

j 13.

Page numbering: Numbering of pages is on a section or for Chapter 3, a specification basis. Consequently and consistent with file management, printing is on a Section basis, with each Section starting on a new page.

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REASON: SGML doe not allow for consecutive page numbering from one file j

to the next. At least not on a practical file management basis.

14.

Figures associated with a Section appear following that Section and not at the end of the Chapter, e.g., the safety limit figure 2.1.1-1 that y

currently follows Section 2.1.2 in the STS.

I REASON: The structure of the document is specified by the DTD and j

things are kept in order.

i 15.

Brackets:

Items that are typical are tagged as such and style sheets l

are formatted to show these items in brackets.

(Typical items are shown within typical tags, blank items, e.g., "[

]", are tagged as such.

Style sheets output the brackets for displaying the typical and blank items.) Brackets are not nested and other tagging schemes are not used, such as option tags, to make any further distinction of degrees of j

typicality or designate any other intent on usage.

1 16.

Where numbers and associated text are extracted from tables, typically 1

instrumentation tables showing functional units, and used in a pseudo lists within the bases, the content of the text (item description) is maintained identical to that in the table to facilitate hyper linking i

4 based on content.

i REASON Maintain consistent content of term content for hyper links.

17.

Bases references to specifications, notably Safety Limits, Actions, Surveillance Requirements. These references are place under an underlined heading of the item, e.g., 6.1 or SR 3.1.5.1 Currently, the j

STS use multiple entries under these heading such as:

A.I. A.2. and A.3 These items are tag references, that hyper link to their source l

specification. Placing multiple items in such references is difficult i

in SGML since styles sheets would have to output the commas and "and" i

used to list multiple items. Because these are in underline format, I

this conflicts with underlining used to indicate hyper links for the i

visually impaired that can distinguish hyper links based on color.

This, in these car.es an icon would be used to create hyper links to specification. Overall, the multiple items create major SGML problems.

l Therefore, this practice will not be allowed by the DTD, and each item l

will require a separate entry.

For now, the SGML instances will i

duplicate the information under each item for current multiple item i

entries, and a Reviewer's Note will be added before the text to indicate such, and that conversions may edit out any materini not specific to in j

the indicated item. Any changes in the text of these items for the SGML

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STS should follow the process for changes in STS in WP51.

i REASON: Current format is incompatible with hyper linking and styling j

of output forms of the SGML instances.

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