Text

Transmits Addl Input to System 80+ Design Control Document
	ML20078R894
Person / Time
Site:	05200002
Issue date:	12/15/1994
From:	Brinkman C; ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	LD-94-071, LD-94-71, NUDOCS 9412280001
	Download: ML20078R894 (504)
	v • d • e

.. _ ______ -

ABB Decembe" 16, 1994 LD-94-07)

Docket No.52-002 Attn: Document Control Desk U.S. Nuclear Regulatory Corr. mission Washington, D.C. 20555

Subject:

System 80+" Design Control Document

Dear Sirs:

This letter transmits additional input to the System 80+ Design Control Document (DCD). Included are the Introduction to the DCD and various markups, all of which have been discussed with the staff.

If you have any questions, please call me or Mr. Stan Ritterbusr,h at (203) 285-5206.

Very truly yours, COMBUSTION ENGINEERING, INC.

C. B. Brinkman Director Nuclear Systems Licensing CBB/ser cc: S. Magruder (NRC)

P. Lang (D0E) n ABB Combustion Engineering Nuclear Power combusion Emmenm. w. N

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P os ett Hill Fid 9412280001 941215 w,noso. ci oces fae ( 28 5 PDR ADOCK 05200002 A PDR

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& 80+ Deakm C adDocument nesism comer.1 noca e Intrai. cts 1.0 Scope and Purpose of the Design Control Document his Desip Control Dm>=aat [DCD) is a repository ofinformation comprising the System 80 +M Standard Plant Desip. De DCD also provides that design-related information to be incorporated by reference in the design corrifimtion rule for the System 80+ Standard Plant Design.

Further sections of this introduction describe the contents and uses of the DCD. De Desip Control Document contains this IntraAWi aa the Certified Design Material (i.e., ' Tier 1"], and the Approved Design Material [i.e., ' Tier 2"] for the System 80+ Standard Plant Design.

1.1 CeitiSed Design Material The Certified Design Material [CDM) for the System 80+ Standard Plant Design includes the following information:

Definitions and General Provisions; e Design Descriptions; Inap~*ia==. Tests, Analyses, and Acceptance Criteria [ITAAC];

Significant Interface Requirements for interfaces t,encs the System 80+ Standard Plant Desip and systems that are wholly or partially outside the scope of the Standard Plant Design; and e Significant Site Parameters.

For ease of reference, the Certified Desip Materialincludes a Table of Contents, a Figure legend, and an Abbreviation List.

1.2 Approved Design Material He Approved Desip Material [ADM), to the extent applicable for the System 80+ Standard Plant Desip, includes:

The information required for the final safety analysis report under 10 CFR 50.34(b);

t System 80+ is a trademark of Combustion Enginecr*mg, Inc.

DCo insoducson Page 1

Swatam 80+ Damian CeedDocenent e Other relevant information required by 10 GR 52.47(a), such as:

- information related to the nree Mile Island requirements under 10 CFR 50.34(f);

- =1 resolutions of the Unresolved Safety Issues, and medium- and high-priority Generic Safety Issues;

- an espanded set of interface requirements and site parameters; and

- important design features identified in risk assessments for the System 80+ Standard Plant Design; and e Emergency Operations Guidelines.

For ease of reference, the ADM contains a general Table of Contents, as well as a specific Table of Contents for each chapter.

L3 R*Intianship of CertiSed Deslan Material to Appmved Design Material ne Design Descriptions, Interface Requirements, and Site Parameters in the CDM are derived entirely from the god.;ans of the Approved Design Material, but may be more general than the design provisions in the ADM. Canplianca with the more detailed Approved Design Material provides a sufficient nethod, but not the only acceptable method, for complying with the more general design provisions in the CDM. However, the methods and provisions specified in the ADM shall be followed imlaan a change is made in accordance with the change processes yWiad in the design certification mie for the System 80+ Standard Plant Design.

L4 Uses of the Design Control Document 1.4.1 Applicability under 10 CFR Part 52 ne design certification rule for the System 80+ Standard Plant Design can be referenced in an application for a Combined maaaa [ COL) under 10 CFR Part 52, Subpart C, and in a subsequently issued COI, Because the DCD is iscwyerated by reference in the design certification rule for the System 80+ Standaal Plant Design, the provisions of the DCD are effective with respect to an application or license that references that rule, with certain exceptions as provided in the mie and described in this Introduction.

1.4.2 General Use De Design Control Document deswibes structures, systems, and components within the scope of the System 80+ Standard Plant Design, including === inted programmatic pw4sions as specified in this document, and the requirements governing the interfaces between the System 80+ Standard Plant Design and plant-specific design feantres. An application for a COL that references the design certification rule for the System 80+ Standard Plant Design must provide a plant-specific Safety Analysis Report [SAR] which shall in::lude information about that part of the plant that is outside the scope of the System 80+ Studard Plant Design or which is otherwise required by a relevant provision of 10 CFR Part 52, but is not included in the DCD. Proprietary references or their equivalent, provided in the application for design certification but not included in the DCD, must be either referenced by or included in the COL Application. Together, the Design Control Document and the plant-specific SAR provide. the technically-relevant information required for a DOD hechaem Ngo a

Swaiam 80+

Denian cann!W2.:

COL, or for an application for a COL, that references the design certi6 cation rule for the System 80+ Standard Plant Design.

2.0 E8ect of the Certi8ed Design Material He following provisions describe the scope and effeet of the Certified Design Material

, 2.1 Conspliance with Certised Design Material All of the information in the Certified Design Material is approved by the NRC, and is applicable to a license application or a Haanse that references the design certification rule for the System 80+

Standard Plant Design; and is among the " matters resolved" under 10 CFR 52.63(a)(4). He provisions and methods specified in the CDM shall be complied with unless a plant specific

exemption is granted by the NRC or a change is made to the CDM in accordance with the change process phi in the design Miscation rule for the System 80+ Standard Plant Design.

2.2 Design Descriptions ne Design Descriptions pennin only to the design of the structures, systems and components of a System 80+ Standard Plant Design and not to their operation, maintenance and administration.

In the event of an inconsistency between the Design Descriptions and the Approved Design Material, the Design Descriptions shall govern.

2.3 Inspections, Tests, Analyses and Acceptance Cdteria An applicant for or holder of a COL shall perform and demonstrate conformance with the ITAAC prior to fuel load. An applicant for a COL may proceed at its own risk with design and procurement activities, and a holder of a COL may proceed at its own risk with design, procurement, construction and preoperational activities, even though the NRC staff may not yet have determined that any particular ITAAC have been satisfied. In the event of a noncomplimee with an ITAAC, the applicant for or holder of a COL shall either take s.orrective actions to comply with that ITAAC or request a change in or cremption from the ITAAC in accordance with the design certification rule for the System 80+ Standard *lant Design.

In accordance with 10 CFR 52.103(g), the Commission must find that the acceptance criteria in the ITAAC are met prior to operation. After the Commission has made the finding required by Section 52.103(g), the ITAAC do not constitute regulatory requirements for subsequent modifications. However, subsequent modifications must comply with the Tier 1 design descriptions, unless changes are made in the Tier 1 design descriptions in accordance with the change processes in the design certification rule for the System 80+ Standard Plant Design. Furthermore, after the NRC has issued its finding in accordance with 10 CFR 52.103(g), the ITAAC do not, by virtue of their inclusion in the Design Control Document, constitute regulatory requirements for the COL holder or for renewals of the COL DCD hoodmeon Page 3

__N _

2.4 Certified Design Material Interface Requirements ne CDM Interface Requirements describe the significant design provisions for interfaces between the System 80+ Standard Plant Design and structures, systems and components that are wholly or partially outside the scope of the System 80 + Standard Plant Design. CDM Interface Requirements also define the significant attributes and performance characteristics that the out-of scope portion of the plant must have in order to support the in-scope portion of the design. He plant-specific SAR shall contain provisions which implement the Interface Requirements in accordance with 10 CFR 52.79(b). Any plant-specific application for a COL shall contain additional ITAAC corresponding to these implementing provisions. In the event of an inconsistency between the CDM Interface Requirements and the Approved Design Material, the CDM Interface Requirements shall govern.

2.5 Site Parameters Site Parameters are phd in the CDM to establish +he bounding parameters to be used in the selection of a suitable site for the facility referencing the System 80+ cenified design. Since the CDM Site Parameters were used in the bounding evaluations of the certified design, they define the requirements for the design that must be met to ensure that a facility built on the site remains in conformance with the design cvsc. tion. In the event of an inconsistency between the CDM Site Parameters and the Approved Design Material, the CDM Site Parameters shall govern.

3.0 EfBect of the Approved Design Material

%e following provisions describe the scope and effect of the Approved Design Material [ADM).

3.1 Compliance with the Approved Design Material All of the information in the ADM is approved by the NRC and, with the exceptions noted in Sections 3.2 and 3.4 below, is applicat.!e to a license application or license that references the desigu certification rule for the System 80+ Standard Plant Design, and is among the " matters resolved" under 10 CFR 52.63(a)(4). Compliance with the Approved Design Materialis a sufficient, but not necessarily the only, method for complying with the CDM. %e provisions and methods specified in the ADM shall be followed unless a change is made in accordance with the change process specified in the design certification rule for the System 80+ Standard Plant Design.

3.2 COL Inibreation items ne Approved Design Material identifies certain matters that need to be addressed by a COL applicant or ba=*e referencing the design certification rule for the System 80+ Standard Plant Design. Dese matters are designated as ' COL Information Items." The purpose of these COL Information Items is to identify the type of information that must be addressed in plant-specific SARs that reference the design certification rule for the System 80+ Standard Plant Design. These COL Licanca Information Items do not establish requirements; rather, they identify an acceptable nco sammunsan Pese 4

SVatun 80+

Deakm Ca' nn~- .,;

set of information, but not the only acceptable set ofinformation, for inclusion in a plant specific SAR. An applicant may deviate from or omit these COL T Wasa Information Items provided that the deviation or omission is identified and justified in the plant-specific SAR. After issuance of a license, the COL T imma Information Items have no further effect for that licensee; instead, the corresponding provisions in the plant specific SAR become applicable. A summary listing of the COL Information Items is provided in Table 1.10-1 of the Approved Design Material 3.3 Approved Design Material Interikce Requirements ne ADM Interface Requirements describe the design provisions for interfaces Mi.a the System 80+ Standard Plant Design and structures, systems and components that are wholly or partially outside the scope of the System 80+ Standard Plant Design. ADM Interface Requirementa, summarized in Table 1.9-1 of the Approved Design Material, also define the attributes and performance characteristics that the out of-scope portion of the plant must have in order to support the in-scope portion of the design. The plant-specific SAR shall contain provisions which implement the ADM Interface Requirements in accordance with 10 CFR 52.79(b). In the event of an inconsistency Mi a.a :he Certified Design Material Interface Requirements and the Approved Design Material Interface Requirements, the CDM Interface Requirements shall govern.

3.4 Conceptnal Designs Conceptual designs for those portions of the plant that are outside the scope of the System 80+

Standard Plant Design are described and designated as out-of-scope in various places in the Approved Design Material As provided by 10 CFR 52.47(a)(1)(ix), these conceptual designs are not a part of the design certification for the System 80+ Standard Plant Design, and do not impose requirements applicable to a COL, nor to an application for a COL, that references the design certification rule.

Textual material comprising Conceptual Design information is denoted by brackets surrounding such material; a listing of this information is provided in Table 1-1.

t 3.5 Plant-Specific Changes to Designated Material la the Approved Design Material Certain information [ Tier 2*) within sections of the Approved Design Material, summarized in Table 1-2, is designated with irm%M text in the ADM. Plant specific changes to any of this italicized design information shall require prior NRC Staff approval. He requirement for prior NRC Staff approval will expire for some of the designated information, as indicated in Table 1-2, when the COL holder first achieves 100% power operation.

3.6 Treatment of Probabilistic Risk Asseanment Information A design-specific Probabilistic Risk Assessment [PRA] for the System 80+ Standard Plant Design was submitted as part of the application for design certification, as required by 10 CFR 52.47. One purpose of the PRA was to develop insights for the design and its features. Significant insights that resulted from the PRA are identified in ADM Section 19.15. However, the detailed methodology and quantitative portions of the design-specific PRA were not included in the DCD because it is I anticipated that this material will be subject to modifications and refinements as the detailed design DCD boodunson ph0e s

System 80+

oeekm caract Document is completed and the as-built plant parameters and new methodology become available.

3.7 Tmatment of Seven Accident Evaluations Chapter 19.11 of the ADM contains various deterministic evaluations of severe accidents for the System 80+ Standard Plant Design. With respect to these evaluations only; a proposed change in the facility or procedures described in the ADM, or a propwd on-site test or on-site experiment, shall be deemed to involve an unreviewed safety question if, as a result of the proposed change, test or experiment:

ne probability of a severe accident previously evaluated in Chapter 19.11 and deemed to be not nedible, increases to the extent that the severe accident is deemed credible; or i

Re postulated consequences to the public of a severe accident previously evaluated in Chapter 19.11 substantially increase.

DCD N Pese s

W 80+ Dee m C 4, i n -;. a Table 11: Index of Conceptual Design Infbr==tian Comesptual Design leau Reference ADM Section Arhninistration Building 1.2.1.4.1.1 Permannel Access Portal 11L4.1.2 W.rh 111.4.13 Switchyard 811.2 Compressed Gas Systems 1 1 11.18; 9.5.10.2; 9.5.10 1 1; 9.5.10.5 Offsite Power Systema 8.1.1; 8.2.1.1; 8.2.1.6 Statian Service Water Panp Struaure 3.8.4.13; 911.2.1.2; 9.2.111.4 Service Water Panp Structure Ventilarian System 9.4.8.2; 9.4.8.5 Ukimate Heat Sink, including SSWS Intake / Discharge 9.2.5.2; 915.4; 9.2.1.5 Potable and Sanitary Water System 9.2.4; 914.2; 9.2.4.2.1; 9.2.412; 9.2.4.5 Offsite Cassmuntentiana 9.512.5 CMa-~ Circulating Water System 10.4.5 1 1; 10.4.5.2; 10.4.5.5 layoutand Fedm! for the laboratory F"Hda- 1333.4.1 Emngency Operations Facility 1333.2.1; 133312 Layout and E? d,=>at for the Onsite Deraataminatian 1333.6.1 oco ==*===a ps 7

l System 80+ Damian Cw ;hn Table 1-2:

h Inder of ADM Items Requiring NRC Approvaglange Items [ Durmelos\ Rafnunce ASME Boiler & Pressure Vessel Code, Secdon HI [ First Pd Power k Table 13 AISC-N690 and ACI-349 Industrial Codes [ Pirst Full Power hies 1-4,15 Design, Quali&=tian and Pi+d ;! Testing for Motor- f Pirst Fd Power ITable 1-6 Operated Valves l

Equipement Seismic Quahharian Methods f First Fd Power Table 1-7 Piping Design Acceptance Criteria f None fTable 18 First Cycle Puel and Control Rod Design First Pull Power / Table 1-9 Instrmaantatian & Controls Setpoint Methodology First Full Power [ Table 110 Instrumentahan & Controls Hardware and Software nar First Full Powerj Table 111 Instrumertation & Controls Enviran=aatal Onehficariaa k First Full Pow [ Table 112 Control Roosn Human Factors 87-Q k None[ Table 113 Note: The applicable portion of the designated Tier 2 reference mat 4 in Tables 1-3 through 1-13 is shown italvi-d within the identdied Approved Design Material [ADM) text or table.

Table 1-3: ASME Boiler & Pressure Vessel Cada.Section III C--*==* ADM Reference ASME Boiler and Pressure Vessel Code, Section IH, Rules for Construction of Table 1.8-6 Nuclear Power Plant Components, Drvision I, Division H ASME Boiler and Pressure Vessel Code, Section HI, Division 1, Subsection NE, 3.8.2.2

" Class MC Compoecats" Table 1-4: AISC-N690 Industrial Code Comunisment ADM Radrence AISC N690, Specifkation for the Design, Fabrication, and Erection of Steel Safety-Related Structores for Nuclear Farihties Table 1.8-6 Analysis and Design of Seismic Category I Steel Structures 3.8.4.5.2 Table 15: ACI-349 Indostrial Code Coasmituneet ADM Reference ACI-349, Code Requirements for Nudear Safety-Related Concrete Structures Table 1.8-6 Analysis and design of Seismic Category I Concrete Structures 3.8.4.5.1 DCD "

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i ra==atment ADM Itaference Design and quahficatian requirements for motor-operated valves 3.9.6.2.L1 i Pre-c f . ! testing of safety-related motor operated valves 33.611.2

! Table 1-7: Equipament Seisade Q=allfleatta= Methods I

s'a = =ne= mat

( ADM Iteference l Seismic quahficaria= reqmrements for =danieml and eledrical equipment 3.10.1.1 M~*ia= of quahficatian method 3.10.L2 Methods and procedures for quablying Seismic Category I electrical equipment 3.10.2.1 and instrumantarian Methods and procedures for quahfying Seismic Category I machmaical 3.10 3.1 equipment including motors Table 1-8: Piping Design Acceptance Criteria eman=ne===t ADM Reference ASME Code and code cases for System 80+ piping and pipe support design L8; App 3.9A [1.1]

Analysis =atkada experimiental stress analysis,indepandant support motion, 3.73.9;3.73.8;3.73.12; inelastu analyss, smaE-bore piping, non-seismic /saiamic interaction, buried 3.73.13;33.13; piping App 3SA [L1]

Piping modehng; piping had---k program, daw : criteria 3.9.1.2.1; App 3SA

[L5.2.2]

Pipe stress analysis criteria; loading and load conabinations, damping values, 3.7.2.15;3.93.1;3S33; combinariam of modal responses, high hywf modes, thermal ancillatians in 333.L43; App 3.9A piping ca==~'*A t o the reactor coolant system, thermal stratification, safety- [L43.2.13, L43.2.L4, related valve design, inat=Ilating and testing, inartianal capability, conibination L43.2.L5,1.43.2.L6, of inertial and seismic motion effects, welded attachments, modal damping for L4.2,1.4.7,1.5.2.2, composite struaures, minimum temperatv:e for thermal analyses. L6.5]; 'Ibl 3.71 Pipe support criteria; appbcable codes, jurisdictianal boundaries, pipe support 3.93.4; App 3SA baseplate and anchar bolt design, use of energy absorbers and limit stops, pipe [L10.1,1.10.2, L7.23, support =M=a==_. seismic self weight ad=+1a= design of supplementary steel, L7.2.8, L7.2.9, L7.2.10, consideration of friction forces, pipe support gaps and clearances, 1.7.4, L7.5) instrianaatation line support criteria Table 19: First Cycle Feet and Contml Rod Design Ceaunitment ADM Reference Fuel and initial core design description and permissible changes 4.1.1 Design features and acceptance criteria for fuel and initial core design Tables 4.1-1,4.12 000 bendusman Pt08 9

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1 Table 110: Inste====*=*1a= & Controls Setpolat Methodology i

! ra==k=.=* ADM Reference J

i Generation of safety system setpoints 7.1.221

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Table 111: Instrussentation & Controls Hardwam and Software Changes t

I Coasmitment ADM Reference i

1 Design, #=?- =,impimaantarian and vahalatian of computer systems 7.1.232

! software changes in safety-related systems i

Table 112: Instr ====* melan & Controls Envirnnunental Qaallfleatlan 1 ca- -k-.=* ADM Reference I

j Eny'ronmental quahficatina of electrical equipment 3.11.1 l

1 j Table 113: Control Roma Human Factom Engineering j thh===* ADM Refenace 1

Human Factors program plan 18.4.2 l Human Factors M==i% verificariam and vahdation plan 18.4.9 Feian=1 task analysis, workload & environmental assumptions and h== 18.5.1.1

{ Task &=---r--a6 and data framework 18.5.13;18.5.13.2; 18.5.133 i Workload loading criteria 18.5.1.4 j Nupier 80+ control room fi=rtianal ask t analysis, scope, PRA and critical 18.5.1.5.1; 18.5.1.5.2; j tasks, information and control requirements, time profile / workload 18.5.1.5 3; 18.5.1.5.4; j ev '->6 Watificarian of overload situatiaan 18.5.1.5.5; 18.5.1.5.6 a

{ Main control room annunciator, display and control imv::Aory 18.5.4 Control room staffing assenptions 18.6.2.2

) Control room console panel profiles 18.6.5.7

! Nuplex 80+ information presentation, standard features i 18.7.1 f Nuplex 80+ safety-related information 18.7.1.8.1 f Remote shutdown panel safety. grade instranentation and controls 18.8.1.1 i

DCD hoodwson Paos 10 4

1 . . _ _ . _ _ .

Syntem 80+ oenka conted ooeummut 5.0 Site Parameters j This section provides a definition of the bounding site parameters used as the basis for the Certified Design, and to be used in evaluating the acceptability of a specific site.

Piping and components of the Certified Design may be designed for site-specific seismic requirements which correspond to a SSE msximum ground acceleration of not less than 0.30g and site soil conditions and properties, j

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Table 5.0-1 Site Parameters Maximum Ground Water Level 2 feet below fmished plant grade level Maximum Mood (or Tsunami) Level 1 foot below finished plant grade level Precipitation (for Roof Design)

Probable Maxinmm Precipitation (PMP) 19.4 inches per hour with a ratio of 0.32 for 5 Esumate (Mammum Average Value Over minute to I hour PMP estimate. (6.2 inches per 5 One Square Mile Area for one hour) minutes)

Maximum Snov land 50 pounds per square foot Design Ambient Temperatures 0% Furmtance Values (Historical Limit N1ndine Peaks < 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months )

Mammum 115'F dry bulb 80*F coincident wet bulb temperature 81*F wet bulb (non-coincident) temperature Minimum 40+p l Extreme Wind l

Basic Wind Speed 110 miles per hour (50 year recurrence) 122 meas per hour (100 year recurrence)

Tornado Muimum Tornado Wind Speed 330 miles per hour Maxunum Pressure Differential 2.4 pounds per square inch Soll Properties Mimmum Static Beanng Capacity 12,000 pounds per square foot at foundation level of Nuclear Island Structure Best Estimate Minimum Shear Wave 700 feet per second l Velocity l

l Liquefaction The soils under safety-related structures and buried piping are stable against liquefaction at the site-specific Safe Shutdown Eanhquake (SSE) level.

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System 80+ contaru contmlDocument Table 5.0-1 Site Parameters (Continued)

I j Seismology ;

SSE Response Spectra Rock Sites See Figures 5.0-1 and 5.0-2.

I Soil Sites See Figures 5.0-3 and 5.0-4. ;

Dilution Factore Dilution Factor T'une Period (sec/m') !

EAB 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.00 x 1&3 LPZ 0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.35 x 104 LPZ 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.00 x 104 LPZ 1 - 4 days 5.40 x 105 LPZ 4 - 30 days 2.20 x 103 l

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A VDU monitor is provided at the Remote Shutdown Panel to provide operational display pages. This

VDU is the same as the control room VDUs and is provided for convenience.

} Local controls, RSP controls, and instrumentation are provided to bring the plant to cold shutdown

conditions utilizing suitable procedures.

$ 1.2.6.3 Discrete Indication and Alann System l The alarm and display systems are designed to aid the operator in handling any challenges to critical plant i availability or safety functions.

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The design integrates the information displayed from alarm windows, meters and VDUs such that the same instruments used for accident monitoring are used for normal plant operations to enable operators

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i to use instruments with which they are most familiar during accident situations.

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The advanced control panels include displays and alarms which allow monitoring of the following criti safety functions:

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Reactivity Control e RCS Inventory Control ( t e RCS Pressure Control g

j e Core Heat Removal /

e RCS Heat Removal e Conniament Integrity

Plant Radiation F2nissions

! The set of VDU panel displays include human engineered pictorial mimic and alarm information that

] provides the operator with a continuous real-time high level overview of the entire power plant's steam j and electrical production process.

i i The alarm and display systems are designed such that no slagle failure will result in the loss of plant information presented to the operator. The design includes diverse means of providing the operator j information necessary to keep the plant operating and for monitoring during accident conditions.

! The alarms are designed to identify their priority through the use of hierarchical physical location and j color coding. Alarm processing techniques (plant mode adaptation and suppression) based on validated i process parameter inputs are used to increase operator comprehension and reduce nuisance alarms.

3 j 1.2.6.4 Data Processing Systesn i

i The Data Processing System (DPS) is a fault tolerant multiprocessor computer based system which

.l provides plant data and status information to the operations staff. The DPS monitors the steam and electrical production processes. It provides the plant operations staff the ability to obtain detailed process

{ data via VDU information output devices.

1

.%....-:Mi n00seniel knreduceion tege 1.212 l

i gy-l Sf m $3 h Vc r. w c .S * ~ U^

System 80+ Design ControlDocument l

[

4. Human factors (i.e., the man-machine interface) are considered throughout the plant and especially in the control room (Chapter 18).

! I

5. ALARA considerations affect the selection of materials and location of piping and equipment that carry radioactive coolant. For example, specifications for the reactor coolant system materials have been :

I tightened to minimize transport of contamination. Improvements in the steam generator tubing material and access openings greatly reduce radiation exposures for maintenance, testing, and inspection. The overall goal is to maintain personnel exposure to less than 100 man-rems per year for each reactor (Chapter 12).

6. Plant security (i.e., sabotage protection) and fire protection concerns have been directly addressed in determining layouts for plant safety systems (Section 13.6).

i l Increased RCS Design Margins and Improvements

i

1. Reactor: The core operating margin has been increased by reducing the normal operating hot leg temperature and revising core parameter monitoring methods. The ability to change operating power level l (i.e., maneuver) using control rods only (without adjusting boron concentration in the coolant system) has been provided, simplifying reactivity control during plant load changes and reducing liquid waste processing l requirements (Sections 4.3 and 4.4).

2. Reactor Pressure Vessel: The reactor vessel is ring-forged with material specifications that result in a sixty year end-of-life RTm well below the current NRC screening criteria. - This results in a significant l reduction in the number of welds (with resulting reduction in inservice inspection) and eliminates concern for pressurized thermal shock (Section 5.3).

3. Pressurizer: The pressurizer volume is increased to enhance the transient response of the RCS and to j reduce unnecessary challenges to safety systems (Section 5.4.g 4 g

4. Steam Generators: The steam generat rs ludeI 690 tubes, improved steam dryers, a seventeen ,

I percent increase in overall heat transfer area @d[ ten percent margin for potential tube plugging. The t j steam generators have a twenty-five percent 1arger secondary feedwater inventory to extend the " boil dry" l

time and improve response to upset conditions. Steam generator improvements also have been added to l facilitate maintenance and long term integrity. 'These include larger and repositioned manways, a standby recirculation nozzle, and a redesigned flow distribution plate (Section 5.4.2).

5. Mechanical improvements based on System 80 startup and operating experience include strengthenM reactor coolant pump impellers, redesigned reactor coolant temperature detector thermowells, strengthened reactor vessel upper guide structure, specification of antimony-free reactor coolant pump bearings, strengthened reactor coolant pump shaf:s, and redesigned steam generator economizer internals.

l Advanced Control Room Design f

1. The Advanced Control Complex (Nuplex 80+) for System 80+ has been designed to meet demanding human factor, reliability, and licensing requirements, and is characterized by state-of-the-art advances, such as distributed digital processing, fiber optic data communications, and touch sensitive video displays (Chapter 18).

2. Nuplex 80 + is a total integration of plant-wide instrumentation and controls (I&C) systems. The Advanced Control Complex includes the Main Control Room, the Technical Support Center, the Remote Shutdown

, Room, Computer Room; the Vital Instrumentation and Equipment Rooms. Non-Essential Electrical l

Equipment Rooms and their respective control, protection, and monitoring systems.

L ,

1 Approved Design Matenal-Introduction Pege dei.,

6cirEP LaAS (Escf ?

System 80+ Design Control Document tJSSS Table 1.4-1 C-EgPressurized Water Reactor Plants Commercial Nominal Plant & Operator Utility Plant Location Operation Mfe Net Non-System 80 Plants t .- v ( .v (._4 t LY ,

Palisades Consumers Power Co. Michigan 197/i 800 Maine Yankee Maine Yankee Atomic Power Maine 1972 No 406199 Co.

Fon Calhoun Omaha Public Power District Nebraska 197JY O 493 Calvert Cliffs % Bal Maryland . 197#5 850 ,

'ves co timore Gas & Electric I

gy z - . . -

c s ny7 ,

m,--

Millstone Point Unit 2 Northeast Utilities Connecticut 1975 865

\ \

/

C- falveynffs Umi 2 0.1;Imcre C= 1 Eicc: ric W ryland _19767 850 3

! Co.

i i

St. Lucie: Unit 1 e Florida Power & Light Co. Florida 1976. ~'

SVo EW %

Arkansas Nuclear One '

Arkansas Power & Light Co. Arkansas 1980 900 Unit 2 ]

/ j

[ --

St-Lucie Unit [) Fiend. Fvwa & Ligm Cu. Fionda g @M

~~

San Onofre. Unit 2 Southern California Edison California r- 1983 - 1100 Co. -

g San Onofre Unit 3'; _ Southern {alifornia me

{alifemia- -1984 - 1100 ~

Cr l' y Waterford Unit 3 Louisiana Power & Light Louisiana 1985 1100 Co.

}

b System 80 Plants ( ( b b ( b f* _

Palo Verde Nuclear Arizona Public Service Arizona Generating Station Company Unit 1 1986 1300 i Unit 2 1986 1300 Unit 3 1988 1300 Washington Nuclear Project Washington Public Power Washington Unit 3 Supply System Deferred 1300 Yonggwang Korea Electric Power Republic of Unit 3 Company Korea 1995* 1000 Unit 4 1996* 1000 Ulchin Korea Electric Power Republic of Unit 3 Company Korea 1998* 1000 Unit 4 1999* 1000 Anticipated Commerical Operation Approved Design Material- Introduction page 1.4 4

r;

( ~

j SF#fam 80+ DesAnn correst Doewnerrt i

1.8 Regulatory C=pHance, Industry Codes and Standards System 80+ compliance with U.S. NRC Regulatory Guides, Generic Letters, Bulletins, and elamanen of i the Standard Review Plan is hh in this section. Regulatory Guides, the guide date or revision, and System 80+ compliance with applicable Guides are summarized in Table 1.8-1.

j Operational experience information highlighted in Regulatory Bulletins and Letters has been incorporated j into the System 80+ design. Generic Letters and NRC Bulletins from 1980 through December,1993 i are identified in Tables 1.8-2 and 1.8-3. 'Ihe applicability of each Generic Letter or Bulletin to System 80+ is assessed, with additional information for applicable issues provided in the referenced !

l sections of this report. I i l System 80+ devinHana from the U.S. NRC Standard Review Plan, NUREG4800 [ LWR N=, June l

j 1987), are listed in Table 1.8-4. Specific sections are also identified where further details relevant to )

i each SRP deviation are discussed. Site-specific compliance with individual Standard Review Pian sections )

is provided in Table 1.8-5. j

{

\

J Table 1.84 identifies the industrial Codes and Standards, and code editions, invoked for certification of l

j the System 80+ Standard Design. Where a particular structure, system, or component requires a code edition different from that listed in Table 1.84, an explanadan of such difference is provided in the

{ appropriate text. Other Codes and Standards that are utilized but not invoked as essential for design j certification are incorpr.sied into the individual chapters of this Safety Analysis Report. Revisions to

accepted industry codes applied to System 80+ will be evaluated on a case-by-case basis. ((The j applicability of code editions will be confirmed by the Combined Operating License applicant in the site-specific Safety Analysis Report.))I l

j ASME Section III, Division 1 Code Cases applicable to System 80+ are identified in Table 1.8-7.

j Except for N-122-1, these Code Cases are consistent with those identified in Regulatory Guide 1.84, Revision 29, for design and fabrication, or Regulatory Guide 1.85, Revision 29, for materials and testing, j that were in effect on July 31,1993.

J Cross-references to subsections of this r rt discussing Unresolved and Generic Safety Issues, the 'Ihree i MileIsland Rule [10 CFR 50.34 (f) new NRC policy issues (SECY-93487) are provided in Tables

. 1.8-8,1.8-9, and 1.8-10.

i

( ,., NA-Lt uhr ok .cau H& m he A& 5W s f sos- q&% ad pp% agsA & & win k p aa A k- 6 4 g 4.*m k k ,3 f l

1 l

8 COL infjmyytion item; see DCD Introde: yon Sectiop 3.2. , , , g t . mr -rc- a 4tw e y rv 1*Pu'r um " r,

^

n.. AD& A000edel-hinF9duellen Pqe r.e r j M '.: f. "" "' t M Tf h I

1

System 80+ Design ControlDocument i Table 1.8-1 NRC Regulatory Guide Applicability Analysis to System 80+

a i

j Document Title Date Section RG 1.1 - Net Positive Suction Head for Emergency Core Cooling and 11/70 6.3;6.5

Containment Heat Removal System Pumps RG 1.2 - Thermal Shock to Reactor Pressure Vessels - GDC 35 Withdrawn j RG 1.3 - N/A (BWR)

RG 1.4 - Assumptions Used for Evaluating the Potential Radiological Rev. 2 6.3.3.6; 12 i Consequences of a LOCA for Pressurized Water Reactors 6n4 RG 1.5 -

f N/A (BWR) i RG 1.6 - Independence Between Redundant Standby (On-Site) Power 3D1 8.1.4.2 Sources and Between Their Distribution Systems j RG 1.7 - Supplement - Control of Combustible Gas Concentration in Rev. 2 6.2.5; 12 Containment Following LOCA (Jg.]

] 1758 i 5 4 XAe RG 1.8 - Qualification & Training of Personnel for Nuclear Power Plants Rev. 2 4/87 y gg k RG 1.9 - Selection, Design, Qualification, and Testing of Emergency Rev. 3 S.I.4.2 l Diesel Generator Units used as Class IE Onsite Electrical Power 7/93 Systems

] .

l RG 1.10 - Mechanical (Cadweld) Splices in Reinforcing Bars of Withdrawn j Concrete Containments 1

) RG 1.11 - Instrument Lines Penetrating Primary Reactor Containment 3/71 7.1.2.15;

} 6.2.4.1.1

) RG 1.12 - Instrumentation for Earthquakes Rev.1 3.7

{ 4n4 RG 1.13 - Spent Fuel Storage Facility Design Basis Rev. 2 9.1 12/81 i

RG 1.14 - Reactor Coolant Pump Flywheel Integrity Rev.1 5.4.1.1

&5 f w 1 4

RG 1.15 - Testing of Reinforcing Bars for Concrete Structures Withdrawn l

{ RG 1.16 - Reporting of Operating Information Rev.4 Not Applicable 1

8M5 i

j RG 1.17 - Protection Against Industrial Sabotage Withdrawn RG 1.18 - Structural Acceptance Tests for Concrete Primary Reactor Withdrawn Containments 4.

l RG 1.19 - Nondestructive Examination of Primary Containment Welds Withdrawn

! RG 1.20 - Comprehensive Vibration Assessment Program for Reactor Rev. 2 3.9.2.4

} Internals During Preoperational and initial Startup Testing 5/76 I

l i

Approved Design Meterial- Introduction Page 1.8-2 i

. - =-

l System 80+ Design ControlD:cument Table 1.8-7 ASME Section III Code Cases Applicable to System 80+

4 Case / Title N-4-11 /[1337-11] Special Type 403 Modified Forgings or Bars, Class 1 and CS; 7/13/87.

N-60;I) Material for Core Support Structures; 7/27/88. /C N-71-15 Additional Materials for Subsection NF, Classes 1,2,3 and MC Component Suppons Fabricated by Welding 12/16/89.

N-122-1 Evaluation of the Design of Rectangular Cross-Section Attachments on Class-1 Piping; 7/27/92.

N-192-2 Use of Braided Flexible Connectors, Class 2 and 3; 9/17/87.

N-247 Cenified Design Repon Summary for Component Standard Suppon, Class 1,2,3, and MC; 1/21/88.

N 249-10 Additional Materials for Subsection NF, Classes 1,2,3 and MC Component Suppons Fabricated without Welding; 5/06/89.

N-262 Resistance Spot Welding for Structural Use in Component Suppons; 7/28/88.

N-284 Metal Contamment Shell Buckling Design Methods;Section III, Division 1, Class MC; 8/25/80.

M-309-1 Identification of Material for Component Suppons: 7/28/88.

N-313 Alternate Rules for half-Coupling Branch Connections, Class 2; 11/28/86.

N-318-4 Evaluation of the Design of Rectangular Cross Section Attachments on Class 2 or 3 Piping; l 12/11/89.

N-319-1 Evaluation of Stresses in Butt Welded Elbows for Class 1 Piping; 7/24/89.

N-391-1 Evaluation of the Design of Hollow Circular Cross Section "/elded Attachments on Class 1 l

Piping; 7/24/89.

l N-392-1 Evaluation of the Design of Hollow Circular Cross Sectior. Welded Attachments on Class 2

, and 3 Pipings; 12/11/89.

l l N-393 Repair Welding Structural Steel Rolled Shapes and Plates for Components Suppons; 7/30/89.

N-411-1 Alternative Damping Values for Response Spectra Analysis for Class 1,2, and 3 Piping; 2/20/89.

N-420 Linear Energy Absorbing Suppons for Subsection NF, Class 1,2, and 3 Construction; 2/14/88

! N-430 Alternative Requirements for Welding Workmanship and Visual Acceptance Criteria for Class 1,2,3, and MC Linear-type and Standard Suppons; 2/28/89.

N-433 Non-threaded Fasteners for Class 1,2, and 3 Components Piping Supports; 12/16/89.

N-474-1 Design Stress Intensities and Yield Strength Values for UNS NO6690 with a minimum specific Yield Strength of 35 ksi, Class 1 Components; 3/05/90.

N-476 Class 1,2,3, and MC Linear Component Suppons - Design Criteria for Single Angle Members, Subsection NF; 5/06/89.

I Approved Design Material Introduction Page 1.8 44

System 80+ Design ControlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues

.g - % p M cel.i. e - >)

No. Title Section*

3 Setpoint Drift in Instrumentation 7. 2 27:> j, n.,

Pa[4J 14 PWR Pipe Cracks 36:, 7 1 47,

! Ch 20 J l

15 Radiation Effects on Reactor Vessel Supports 5.4.14.2 3 Ch 20 22 Inadvertent Boron Dilution Events 7.7.1.1.10 15.4.6 Ch 20 23 Reactor Coolant Pump Seal Failures 5.4.1.3 8.1.4.2 9.3.4 Ch 20 29 Bolting Degradation or Failures in Nuclear Plants 3.9.3 4.5.2.1 5.2 Ch 20 t 36 Loss of Service Water 9.2.1 l l

Ch 20 43 Reliability of Air Systems 9.3.1 Ch 20 l

l 45 Inoperability of Instruments Due to Extreme Cold Weather 9.4 Ch 16 Ch 20 48 LCO for Class IE Vital Instrument Buses in Operating 8.3.2 I Reactors Ch 16 Ch 20

/

i

' /

\ t/fg i

_7 USIs and GSis not applicable to the System 80+ Standard Plant Design are identified in Chapter 20, along with the corresponding reason.

Approved Design Material Introductas Page 1.8-4S

Syotem 80+ Design C7atrolDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.)

m r OW e

,, fY No. [ Title /

u /

[%/Section[ ut

&%R 49 7nrerlocks and LCOs for Redundant Class IE Tie Breaker 8.3.1.2 g ,,4 Ch 16 Ch 20 Nb cas-51 Proposed Requirements for Improving Reliability of Open Cycle 9.2.1 Service Water Systems 9.2.2 6N I 9.2.5 Ch 20 57 E*fects of Fire Protection System Actuation on Safety Related 9.5.1 Equipment Ch 20 1

I 64 Identification of Protection System Instrument Sensing Lines 7.1.2.31 Ch 20 66 Steam Generator Requirements 5.4.2 10.3.5 10.4.1.2 l

Ch 20 67.3.3 Steam Generator Staff Actions; Improved Accident Monitoring 7.5 Ch 20 70 PORV and Block Valve Reliabilirf 6.7 Ch 20 75 Generic Implications of ATWS Events at Sal;*u 7.7.1.1.11 Ch 20 78 Monitoring of Fatigue Transient Limits for Reactor Coolant System 3.9.1 ,

! Ch 20 I 79 Unanalyzed Reactor Vessel Thermal Stress during Natural Convection 3.9 Cooldown 5.3 Ch 20 82 Beyond Design Bases Accidents in Spent Fuel Pools 9.1.2 9.1.3 9.1.4 i Ch 20 83 Control Room Habitability 6.4 l

9.4.1 Ch 20 87 Failure of HPCI Steam Line Without Isolation (in BWRs) 3.9.6 Ch 20 93 Steam Binding of Auxiliary Feedwater Pumps 10.4.9.5.2 Ch 20 94 Additional LTOP for Light Water Reactors 5.2.2.10 5.2.3 5.3 Ch 20 l

Approved Design Matenal Introduction Pope 1.8-46 l

l System 80+ Design ControlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.) ;

1 l

No. Title Section l 99 Loss of RHR Capability in PWRs 5.4.7 Af) 19.8A i

Ch 20 i 103 Design for Probably Maximum Precipitation 2.0 l

3.1.2 '

Ch 20

105 Interfacing Systems LOCA at LWRs Aff SE l Ch 20 l 106 Piping and Use of Highly Combustible Gases in Vital Areas; Fire 9.5.10 Protection Ch 20 113 Dynamic Qualification and Testing of Large Bore Hydraulic Snubbers 3.9.3.4 Ch 20 118 Tendon Anchorage Failure 3.8 Ch 20 119.1 Pipe Rupture Requirements 3.6.2.1 3.9.2.5 3.9.3.1 Ch 20 l

119.2 Pipe Damping Values 3.7.1.3 l Ch 20 119.3 Decoupling OBE from SSE 2.5 3.7 Ch 20 ,

119.5 leak Detection Requirements 3.6.3.3 5.2.5 7.7.1.6 Ch 20 l i

120 On-Line Testability of Protection Systems Ch 16 Ch 20 121 Hydrogen Control for Large, Dry PWR Containments 3.8 6.2.5 l 19.11 Ch 20 122.2 Initiating Feed and Bleed 7.5.1.1.5 10.4.9 Ch 20 124 Auxiliary Feedwater System Reliability 10.4.9 Ch 20 125.I.3 SPDS Availability 7.5 7.7.1 18.7.1 Ch 20 Approved Design Material Introduction Page 1.8-47

System 80+ Design ControlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.)

i No. Title Section 125.11.7 Reevaluate Provision to Automatically Isolate Feedwater from Steam 10.4.9 j Generator During Line Break Ch 20

. 128 Electrical Power Reliability 8.3

Ch 20 130 Essential Service Water Pump Failure at Multiplant Sites 1.2.1.3 9.2.1 Ch 20

. 135 Integrated Steam Generator issues 5.4.2 6.7.2 3 7.3.1 l 7.5.1 10.3.2

. 10.4

15.6.3 i Ch 20

142 leakage Through Electrical isolators in Instrumentation Circuits Ch 20 143 Availability of Chilled Water Systems 9.2.9 7 9.4 j Ch 20 1 153 less of Essential Service Water in LWRs 9.2.1 l Ch 16 Ch 20 155.1 More Realistic Source Term Assumptions 3.11

}

6.5 Af$15A Ch 20 A-1 Water Hammer 5.4.7 6.3.1

6.5.1 j 7.7.1.1.4

, 9.2

! 10.3.2.2 i 10.4.7 l 10.4.9 i Ch 20 l A-2 Asymmetric Blowdown Loads on RCS 3.6.2 3.6.3 Ch 20 A-4 C-E Steam Generator Tube Integrity 5.4.2 10.3.5

10.3.6 s 10.4.1

! 10.4.6 i

10.4.8 Ch 20 Approved Design Material-Introduction Page 1.8M

Syatem 80+ Design controlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.) >

No. Title Section A-9 Anticipated Transients without SC ATWS) 7 1

/ 7.7.1.1. ! !

/ C. 19.0 Ch 20 l A-12 Fracture Toughness of Steam Generator & RCP Supports 5.2 5.4.14 Ch 20 A-13 Snubber Operability Assurance 3.9.3.4 Ch 20 +

t A-17 Systems Interaction 3.6 l 19.0 Ch 20

)

A-24 Qualification of Class IE Safety Related Equipment 3.9.2.2 3.10 1 3.11 Ch 20 A-25 Non-Safety 1. cads on Class IE Power Sources 7.1.1 8.1.3 l 8.1.4 8.3.1.2.7 Ch 20 A-26 Reactor Vessel Pressure Transient Protection 5.2.2 Aff SA 7.2 Ch 20 A-29 Plant Design for Reduction of Vulnerability to Sabotage Aff13A Ch 20 A-30 Adequacy of Safety Related DC Power Supplies 8.3.2 Ch 20 i

A-31 RHR Shutdown Requirements 5.4.7 6.7 8.1.2 8.2 10.1 10.4.9 A-35 Adequacy of Offsite Power Systems 8.1.3 l 8.2 l 8.3.1 Ch 20 A-36 Control of Heavy Loads Near Spent Fuel ,9,1.4 Ap$19.8A Ch 20 Anoroved Design Meterial- Introduction page 1.849

Syntem 804- _

oesign controlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Lssues (Cont'd.)

No. Title Section A-40 Seismic Design; Short Term Program 2.5 3.7 Ch 20 A-43 Contamment Emergency Sump Performance 6.8 Afp.19.8A Ch 20 A-44 Station Blackout 8.1.4.2 8.3.1.1.5 19.0 Ch 20 A-45 Shutdown Decay Heat Removal Requirements 5.4.7 Ah19.8A Ch 20 i

A-47 Safety implications of Control Systems 6.3.2 7.3.1.1.10.3 7.3.2 10.4.7 10.4.9 Ch 20 A-49 Pressunzed Thermal Shock 5.2.2.11 5.3 Ch 20 i

B-5 Ductility of Two-Way Slabs & Shells: Steel Contamments 3.6 3.8 Ch 20 B-17 Criteria For Safety-Related Operator Actions Ch 7 Ch 20 B-36 Develop Design Testing, and Maintenance Criteria for Atmosphere 9.4 Clean-up System Air Filtration and Adsorption Units for Engineered Ch 20 l Safety Features Systems and for Normal Ventilation Systems

! B-53 Lead Break Switch 8.1 8.2 8.3 Ch 20 B-56 Diesel Generator Reliability 8.1 8.3.1.1.4 Ch 20 B-60 Loose Parts Monitoring System 7.7.1.6.3 Ch 20

! B-61 Allowable ECCS Equipment Outage Periods Ch 16 Ch 20 l

I l

Approved Design Material- Introduction Pope 1.8-50

-,v .- , i

Sy tem 80 + D sign centrolD:cument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.)

No. Title Section ,

i B-63 Isolation of low Pressure System Connected to The Reactor Coolant 3.2.2 l Pressure Boundary 3.9.6.2 l AMSE Ch 20 B-66 Control Room Infiltration Measurements 6.4 9.4.1 Ch 20 C-1 Assurance of Continuous lAng-term Capability of Hermetic Seals on 3.11 Instrumentation and Electrical Equipment Ch 6 Ch 15 19.11.4 Ch 20 C-2 Study of Contamment Depressurization by inadvenent Spray 3.8 Operation 7.3.1 18.3 Ch 20 C-4 Statistical Methods for ECCS Analysis 6.3 Ch 20 C-5 Decay Heat Update 6.2 6.3 15.0 Ch 20 C-10 Effective Operation of Contamment Sprays in a LOCA 6.5 15.6.5 Ch 16 Ch 20 C-12 Pnmary System Vibration Assessment 3.9.2 7.7.16 Ch 20 HF 5.1 local Control Stations 9.5.2 18.7.1 6.2 Ch 20 HF 5.2 Review Criteria for Human Factors Aspects of Advanced Controls 18.3 and Instrumentation 18.4 18.7 Ch 20 1.C.1 (1-4) Shon Term Accident Analysis and Procedures Revision 1.6 Ch 20 1.C.9 Ieng-term Program Plan for Upgrading of Procedures Ch 20 1.D.1 Control Room Design Reviews 18.0 Ch 20 Approved Design Material- Introduction Page 1.8-51

Syntem 80+ Design ControlDocument Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.)

l No. Title Section I.D.2 Control Room Design; Plant Safety Parameter Display Console 7.7.1.4 7.7.1.7 13.3.3 18.7.1 Ch 20 I.D.3 Control Room Design; Safety System Status Monitoring 7.1.2.21 Ch 20 I.D.4 Control Room Design Standard 18.3 i 18.4 I 18.7 1 Ch 20 l I.D.5 (1) Control Room Design; Improved Instrumentation Research - 18.0 Operator / Process Communication Ch 20 1.D.5 (2) Control Room Design; Improved Instrumentation Research - Plant 7.1.2.21 Status and Post-Accident Monitoring 7.2.1.1.5 7.5.1.1.5 18.7.1.8 j Ch 20 i i

1.D.5 (3) Control Room Design; 5.2.5 i On-Line Reactor Surveillance System 7.1.2.20 7.7.1.6 Ch 20 i I.D.5 (4) Control Room Design; improved Instrumentation Research - 5.2.5.1.2.1 Process Monitoring Instrumentation 7.5.1 l 7.5.2.5 Ch 20 1.F.1 Quality Assurance; Expand Quality Assurance List for Equipment 3.2 Imponant to Safety 17.1 Ch 20 I.F.2 (6,9) Quality Assurance; Develop More Detailed QA Criteria 17.1 Ch 20 I.G.2 Scope of Preoperational and lew-Power Testing Program Ch 20 II.B.1 Safety Review Consideration; Reactor Coolant System Vents 6.7.1.2.1 l 6.7.2.1.1 7.5.1 Ch 20 II.B.2 Safety Review Consideration; Plant Shielding to Provide Post 12.2.3 Accident Access to Vital Areas 12.3.1.2 12.3.1.3 Ch 20 II.B.3 Safety Review Consideration; Post Accident Sampling System 9.3.2 Ch 20 Approved Design Material . Introduction Page 1.8-52

I System 80+ Design centrolDocument l

l Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.)

i l

No. Title Section ll.B.8 Rulemakmg Proceedings on Degraded Core Accident; Hydrogen Rule, 19.11 Severe Accident, Etc. Ch 20 II.C.4 Reliability Engineering 17.3 19.15 Ch 20 II.D.1 Coolant System Valves; Testing Requirements 5.4.13.4.1 6.7 7.7.1.1.11 Ch 20 II.D.3 Coolant System Valves; Valve Position Indication 5.2.5.1.2.1 5.4.13 7.7.1.6 Ch 20 ll.E.1.1 Auxiliary Feedwater System Evaluation 10.4.9 Ch 20

!!.E.1.2 Auxiliary Feedwater System Automatic Initiation an Flow Indication 7.3 7.7 10.4.9 Ch 20 II.E.3.1 Decay Heat Removal; Reliability of Power Supplies for Natural AQ SD Circulation 8.3.1.1.2 Ch 20 ll.E.4.1 Containment Design; Dedicated Penetrations 6.2.4 l 6.2.5 Ch 20 ll.E.4.2 Contamment Design; Isolation Dependability 3.1 6.2.4 9.4.6 i Ch 20

!!.E.4.4 (1-5) Contamment Design; Purging 6.2.4 9.4.6 l Ch 16 Ch 20 l II.E.6.1 In-situ Testing of Valves-Test Adequacy Study 3.9.6 Ch 20 II.F.1 Additional Accident Monitoring Instrumentation 7.5.1.1.5 Ch 20 II.F.2 Identification and Recovery from Conditions leading to inadequate 7.5.1.1.7 Core Cooling Ch 20 ll.F.3 Instrumentation for Monitoring Accident Conditions 7.5.1.1.5 Ch 20 l

1 Approved Design Meterial- Introouction page 1.g.53

i i System 80+ Desen CsntrolDocument i

. Table 1.8-8 Cross-Reference for Unresolved and Generic Safety Issues (Cont'd.) '

1 i

No. Title Section ;

11.G.1 Power Supplies fee Pressurizer Relief Valves, Block Valves, and 6.7 Level Indicators 7.5.2.5

$ Ch 20 I

j II.J.3.1 Orgamntion and Staffing to Oversee Design and Construction Ch 20 II.K.1 (1) (1,2,4(A-C),5,7,8,10-13. <- 6.3.3 j 17-23) Measures to Mitigate Small Break LOCAs and Feedwater y Ch 20 j Accidents: NRC Bulletins a-

) II.K.1 (2) (3,44,6,9,14-16,24-28) Measures to Mitigate Smal LOCAs& wA 6.3.3 i Loss ofpV Accidents: NRC Bulletins Ch 20 II.K.37 '([5,6,8,25,30,31,55) Final Recommendations of B&O Task Force to 6.3.3 i Mitigate Accidents 9.3.4.3.2

gg 19.3.3.1 Ch 20 lgy j III. A.1.2 (1-3) Upgrade Licensee Emergency Support Facilities 13.3.3.1 J 13.3.3.2 13.3.3.3 Ch 20 Ill.D. I.1 (2) Review Information on Provisions for Leak Detection 11.5 l 12.1 j 12.3

Ch 20 1

III.D.3.3 (1-4) In. Plant Radiation Monitoring 7.5.1.1.5 i j, 11.5 j 12.3.4 j

Ch 20

=

j Ill.D.3.4 Control Room Habitability 6.4 t

Ch 20 i

i j

l i

4 1

i ApproM Design Meterial kutroducten page 1.g.54

System 80+ Design ControlDocument I

Table 1.8-9 Cross-Reference for the TMI Rule (10 CFR 50.34f) gp'q i FA ^

,- -q No. Title Section*

(1)(i) Plant / Site Specific PRA (II.B.8) 19.15 No o .c. c @ I lei,A l.4 ott (1)(ii)(A) Simplified AFWs Reliability Analysis (II.E.1.1) 10.4.9; 19.6% M a ;. . e n . ,-Ch 20- W' ;7 q Design Review of AFWs 01.E.1.1) '

- (1)(ii)(B) 10.4.9)

,2 o .:. -t e-Ch-20 gg4a (1)(iiXC) Evaluation of AFWs Flow Design Bases and Criteria 10.4.9

&.11..J Rd

,cGla m n (1)(iii) Evaluation of RCP Seal Damage Following Small Break LOCA with 5.4.1.33 ' J#

LOP (II.K.2.16 & II.K.3.25) 8.1.4.2 ,

9.3.4 ',

gg -

/3YYS- :. 2 . t u - +-Ch 20 $ ],

(1)(iv) Analysis of Probability of Small'$caid LOCA Caused by IORV U

$ 9 N *[ g;l' a GI.K.3.2) -,y. r. r2 I - ; Ch /

l IS -

(1)(v) Evaluation of Effectiveness of High Pressure Coolant Injection N/A (BWRs Only) GI.K.3.16)

(1)(vi) Reduction of Challenges to Relief Valves (BWRs Only) (II.K.3.16) N/A f

9 I (1)(vii) Feasibility and Risk Assessment of ADS Design Modifijons N/A (BWRs Only)(II.K.3.18)

(1)(viii) Effect of Core Cooling Modes Under Accident Conditions (BWRs N/A Only) 61.K.3.21)

(1)(ix) Study of Additional Space Cooling Needs for RCIC & HPIC (BWRs N/A V' ,

Only) (II.K.3.24) p $

(1)(x) Study ADS Capability During and Following Accident Conditions (BWRs Only)(II.K.3.28)

N/A[

(1)(xi) Evaluate Alternate Depressurization Methods (BWRs Only) N/A 01.K.3.45) 4.

(1)(xiiXA) Compare Costs and Benefits of Alternative Hydrogen Control 19A.S.9 g Systems ,

(1)(xii)(B) Verify Compliance with (F)(1)Ox) of Selected Hydrogen Control 6.2.5.1.2i

^ ' ' ;.

System I Aff 19.11K .

i (1)(xiiXC) Evaluate Design, Function & Layout of Alternative Hydrogen 6.2.5.1.23

! Control Systems e Afp19.11K (2)(i) Simulator Capability (COL Requirementf)'(I.A.4.2) Ch 20 (2Xii) Pr gram to Improve Procedures (COL Requirement) 0.C.9) Ch 20 - 1 N- %

N/A indicates items which are/not applicable 9to the System 80+ Standard Plant Design.

/ / / \

p -

Approved Design Material- Introduction Pope 1.8-55 l ,_ _ _ _ _ _ _

System 80+ Design Conso! Document l l

l Table 1.8-9 Cross-Reference for the TMI Rule (10 CFR 50.34f) (Cont'd.)

1 No. Title Secdon g

(2Xiii) Control Room Design (l.D.I.) 18.0 l v..9a - ^ Ch% b l (2Xiv) Safety Parameter Display Console (I.D.2) 7.7.1.4 >

7.7.1.7; p

13.3.3 ;

18.7.1*, c- -

7E so.z . 91 =m {M ' .

(2Xv) Indication of Bypassed and Operable Status of Safety

-*Ch-20 7.1.2.21 $ 7' l Systems (l.D.3) w. 2. o 2 -

CE go) . ;l (2Xvi) RCS High Point Venting (II.B.1) 6.7.1.2.1 6.7.2.1.1 7.5.1 0:/. 2 . / o / .__ __.. Ch3 l

(2 Xvii) Radiation and Shielding Design Review (II.B.2) 12.2.3 12.3.1.2 12.3.1.3 w.t. /n > Gr-20 (2)(viii) Post-Accident Sampling System (II.B.3) 9.3.2 v o. z . fo y, , <b-20 (2XixXA) Hydrogen Control System for 100% Clad / Metal-Water 6.2.5 Reaction (II.B.8) - Capability to Maintain < 10% 19.11.4 Att 19.11K (2XixXB) Hydrogen Control System for 100% Clad / Metal - 6.2.5 Water Reaction (11.B.8)- Assure No local Pockets Ag 19.11K That Could Cause less of Containment orMitigating Features !

(2Xix)(C) Hydrogen Control System for 100% Clad / Metal - 19.11.4 Water Reaction (II.B.8) - Equipment Need for Safe Shutdown Qualified for Environment (2)(ix)(D) Hydrogen Control System for 100% Clad / Metal - N/A Water Reaction (II.B.8) - Inadvertent Actuation of inerting System (2)(x) Qualification of RCS Relief, Safety and PORV Block 5.4.13.4.1; Valves (II.D.1) 6.7) 7.7.1.1.113 2 e . . It c - - > -Cb (2Xxi) Indication of Relief and Safety Valve Position in 5.2.5.1.2.1 Control Room (II.D.3) 5.4.13 7.7.1.6 Eh 20 Approved Design Material Introduction Page 1.8 56

System 80+ Design CrntrolDicument J

J .

Table 1.8-9 Cross-Reference for the TMI Rule (10 CFR 50.34f) (Cont'd.) )

i

No. Title Section (2)(xii) Actuation and Indication of AFW Flow in Control 7.3

Room (II.E.1.2) 7.7

10.4.9 l -g.2, /c 9 - Ch 20 (2)(xiii) Cavability to Maintain Natural Circulation and Afh5D Pressurizer Heaters with Onsite Power (II.E.3.1) 8.3.1.1.2 ye. ; . /io - - -Ch40-(2)(xiv)(A) Contamment Isolation System (ll.E.4.2) - Automatic 6.2.4 Isolation on Non-Essential Systems 9.4.6 i

Ch 20 v .2. . /,2-1 I (2)(xiv)(B) Contamment Isolation System (II.E.4.2) - Two 6.2.4 4 Isolation Barriers in Series for Each Non-Essential ;-o. > it >

Penetration (2)(xiv)(C) Containment Isolation System (ll.E.4.2) - Reset of 6.2.4 CIAS Does Not Result in Valves Opemng o c 2 - I' >

(2)(xiv)(D) Contamment Isolation System (II.E.4.2) - CIAS 6.2.4 Setpoint as IAw as Compatible with Normal Operation 2 a ; l' t-f (2)(xiv)(E) Containment Isolation System (II E.4.2) - Automatic 6.2.4 l Isolation of All Paths to "Ihe Environs on High ; v : .It :-

,1 Radiation (2)(xv) Contamment Purging / Venting Capability with Reliable 6.2.4

]

j isolation During Accident (ll.E.4.4) 9.4.6 i TS .9 g Ch 20~ L .. 1 1 (2)(xvi) Establish Design Criterion for Actuations of ECS and N/A

RPS (BWRs Only) (II.E.5.1)

(2)(xvii) Provide Instrumentation to Measure, Record and 7.5.1.1.5 ,

Indicate in the Control Room (II.F.1) Ch 20 . o. '- /> 5 l I (2)(xviii) Control Room Indication of inadequate Core Cooling - 7.5.1.1.7

Saturation Meter (ll.F.2) Ch 20- _ , / , c, (2)(xix) Provide Post-Accident Monitoring 7.5.1.1.5

, Instrumentation (ll.F.3) Gh40 _v.1 . /r7 (2)(xx) Power Supplies for RCS Relief and Block Valves and 6.7 level Indicators (ll.G.1) 7.5.2.5

_Ch 20 0 > t. /' 3 I

(2)(xxi) Auxiliary Heat Removal System Design (BWRs Only) -N/A--

(II.K.1.22) % E,f', '

i j (2)(xxii) FMEA on Integrated Control System (B&W Only) N/A (II.K.2.9) i (2)(xxiii) Anticipatory RPS Trip on less of MFW and Turbine N/A Trip (B&W Only) (II.K.2.10)

Approved Design Material- Introduction Pope 1.8-57

i Syat m 80+ Design C:ntr:/ D:cument l l

Table 1.8-9 Cross-Reference for the TMI Rule (10 CFR 50.34f) (Cont'd.)

No. Title Section (2)(xvii) Provide Instrumentation to Measure, Record and Indicate in the 7.5.1.1.5; Control Room (II.F.1) 20.2.115 (2)(xviii) Control Room Indication of Inadequate Core Cooling - Saturation 7.5.1.1.7; Meter (II.F.2) 20.2.I16 (2)(xix) Provide Post-Accident Monitoring Instrumentation (ll.F.3) 7.5.1.1.5; 20.2.I17 (2)(xx) Power Supplies for RCS Relief and Block Valves and Level 6.7; 7.5.2.5 Indicators (II.G.1) 20.2.118 (2)(xxi) Auxiliary Heat Removal System Design (BWRs Only)(II.K.I.22)

N/A (2)(xxii) FMEA on Integrated Control Sys gB&W Only) (ll.K.2.9) N/A (2)(xxiii) Anticipatory RPS Trip on Loss of MF N and Turbine Trip (B&W N/A l Only) (II.K.2.10) i (2)(xxiv) Recording of Post-Accident Reactor Vessel Water Level (BWRs N/A !

Only) (II.K.3.23)

(2)(xxv) Onsite Technical Support Center, and Technical Operations Center, 13.3.3.1; and Emergency Operations Facility (III.A.I.2) 13.3.3.2; 13.3.3.3; l 20.2.122

! (2)(xxvi) leakage Control and Detection Design and Program for Systems 11.5; 12.1; l l Outside Containment (III.D.I.1) 12.3; 20.2.125 t

(2)(xxvii) Monitoririjof Inplant and Airborne Radioactivity (Ill.D.3.3) 7.5.1.1.5; 11.5;

! 12.3.4; 20.2.126 (2)(xxviii) Evaluate Potential Pathways That May lead to Control Room 6.4; 20.2.127 Habitability Problems Under Accident Conditions (III.D.3.4) s m.

(3)(i) Administrative Procedures for Evaluating Industry Operating, Design, N/A yN (3)(iii and Construction Experience During Design and Construction (I.C.5)

Ensure that QA List Contains All Systems, Structures and 3.2; 17.1; h p;A s

f Components important to Safety Per Criterion 11 of 10 CFR 50 20.2.98 A b' Appendix B (I.F.1) t$ p (3)(iii) Quality Assurance Program (I.F.2) 17.1; 20.2.99 U I (3)(iv) Provision of Dedica'ed Containment Penetrr: ion for Future App 19A; 5.2; 4 ) v Installation of Systems to Prevent Containment Failure (ll.B.8) 20.2.104 (3)(v)(A) Containment Integrity During Hydrogen Burn (or Inerting) for 100% 3.8.2; Clad / Metal - Water Reaction (II.B.8) App 19.llK; 20.2.104 (3)(v)(B) Containment Structural loading from Inadvertent Actuation of N/A Inerting System (II.B 8)

(3)(vi) External Hydrogen Recombiners (II.E.4.1) 6.2.4; 6.2.5; 20.2.I i1 (3)(vii) Management Plan for Design and Construction Activities (COL 20.2.119 Requirement) (ll.J.3.1)

Approved Design Material

Introduction Page 1.8-49

Syst:m 80+ orsign centr t 0:cument Table 1.8-10 Cross-Reference for New NRC Policy Issues (SECY-93-087)

No. Title Section* )

'/

1. A Use of a Physically Based Source Term ).'ll: 6. pp, I.B Anticipated 'iransients Without Scram ( 7.3;7.7M.11 I.C Mid-Loop Operation %p 19.8A 1.D Station Blackout 8.1.4.2:

8.3.1.1.5 1.E Fire Protection 9.5.1 1.F Intersystem Loss-of-Coolant Accident App SE f

I.G Hydrogen Control 6.2.5; 19.11.3 I.H Core Debris Coolability 19.11.3 I.I High-Pressure Core Melt Ejection 19.11.3 r\.

I.J Containment Performance 19.11.3 '.'

?

1.K Dedicated Containment Vent Penetration 19.15.5 g l

1.L Equipment Survivability 19.11.4.4 '

I.M Elimination of Operating-Basis Earthquake 2.5; 3.7 1.N Inservice Testing of Pumps and Valves 3.9.6; 5.2.4; E l

6.6 II.A Industry Codes and Standards 1.8 'r ,

II.B Electrical Distribution 8.2:8.3 II.C Seismic Hazard Curves and Design Parameters 19 7 /

II.D Leak-Before-Break 3.6.2.1.3; IL7b 3.6.3 g g([6 II.E. Classification of Main Steamlines in Boiling Water Reactors N/A II.F Tornado Design Basis 2.3.2.1 II.G Containment Bypass 6.2.2; App SE II.H Containment Leak Rate Testing 3.8.2.7 11.1 Post-Accident Sampling System 9.3.2 II.J 1.evel of Detail 1.1.1; 1.2.2 II.K Prototyping N/A II.L ITAAC 14.3 II.M Reliability Assurance Program 17.3 "N/A" indicates items which are Not Applicable to the 53 m 80+ Standard Plant Design.

Approved Design Material-Introduction Page 1.8-50

1 System 80+ Deslan Controlkwument I

Table 1.8-10 Cross-Reference for New 087) (Cont'd.) NRC Policy %s(

No. Title Section _l4

'..b II.J Level of Detail ; A II.K Prototyping N/A II.L ITAAC 14.3 r II.M Reliability Assurance Program 17.3 II.N. Site-Specific Probabilistic Risk Assessments and 17.3; Analysis of External Events 19.7 19.15 II.O Severe Accident Mitigation Design Alternatives 19.15.5 Aff 19A II.P Generic Rulemahng Related to Design Certification 'N/A II.Q Defense Against Common-Mode Failures in Digital 7.2 .

Instrumentation and Control Systems 7.3 7.7 App?A II.R Steam Generator Tube Ruptures .d9.4.4-- # 9 .19 .7. f O II.S PRA Beyond Design Certification 17.3 19.7 19.15 II.T Control Room Annunciator (Alarm) Reliability 7.7 III.A Regulatory Treatment of Nonsafety Systems in N/A Passive Designs III.B Definition of Passive Failure N/A III.C SBWR Stability (Passive Design) N/A III.D Safe Shutdown Requirements (Passive Design) N/A III.E Control Room Habitability (Passive Design) N/A r i

III.F Radionuclide Attenuation (Passive Design) N/A III.G Simplification of Offsite Emergency Planning 15.6.5 III.H Role of the Passive Plant Control Room Operator N/A Approved Desiprs Material-Introduction Pege r.8-60

System 80+ Design ControlDccument 1.10 System 80+ COL Information The System 80+ Standard Design represents an essentially complete nuclear plant. However, certain topics are more appropriately addressed by a COL / applicant referencing this DCD.

Table 1.10-1 identifies ADM sections where descriptions of COL information items are presented.

Neither the table listings nor the descriptions within the cited ADM sections are intended to constitute requirements for the COL applicant. This information is provided only for purposes of facilitating a COL applicant's preparation of its COL application.

Table 1.10-1 COL License Information COL No. FSER No. Section Subject 1-1 1.8 Applicable editions of industry codes and standards 1-2 1.9 Design interfaces 2-1 2.0-1 2.0 Site Parameters 2-2 2.1-1 2.1 Geography and Demcgraphy Information 2-3 2.2-1 2.2 Industrial, Transportation and Military Hazards 2-4 2.3 2.3 Meteorology Information 2-5 2.4-1 2.4 Hydrologic Information 2-6 2.5-1 2.5 Geology, Seismology and Geotechnical Information 3-1 3.4-1 3.4 Flood Analysis 3-2 3.5-1 3.5 Missile Protection 3-3 3.6-1 3.6 Final Piping Design Information I 3-4 3.7-1 3.7 Site and Plant-Specific Seismic Design Information l 3-5 3.8-1 3.8 Site and Plant-Specific Structural Design Information 3-6 3.9-1 3.9 Site and Plant-Specific Information for Mechanical System and Components Design 3-7 3.10-1 3.10 Seismic and Dynamic Qualification Program Details 3-8 3.11-1 3.11 Environmental Qualification Program Details 4-1 4.2.7-1 4.2.3.2.1 .K Online fuel failure monitoring and post-

/

irradiation surveillance P 4-2 4.4.4-1 7.2.1.1.2.5 CPC/CEAC software testing and change control 5-1 5.2.1.1-1 5.2.1.1 ASME Code edition, Addenda and Code cases for construction of the reactor coolant pressure boundary components Approved Design Meterial- introduction Page 1.10-1

k n.h& hbr C c 'i M 'r: ON

~

_( 0 r. kj my ,j% .H A & L WJ-N+

System 80 + -.&W ,L. Design ControlDocument 4

1 Table 1.10-1 COL License Information (Cont'd.) l 9 c, ,

COL No. FSER No. Section Subject 5-2 5.2.2.2- 1, 5.2.2.3- 1, 5.2.2.10.2.2, Verification of the material properties and end- 's-5.3.1-1, 5.3.2-1 5.2.3.1, 5.3.2, of-life fluence and resulting P-T limits and ,l 5.4.14.3 LTOP temperatures j 5-3 5.2.4-1, 6.6-1 5.2.4,6.6.1 PSI and ISI program plans for NRC staff review b I 5-4 5.4-1 5.4.2.5 Steam Generator tube inservice inspection 9 *-

program @N 5-5 SF-1 5.6.3 (App. 5F) 12akage monitoring program V 6-1 6.1.1 - 1, 6.1.1 -2, 4CM-6-4 +- Engineered Safety Feature Systems materials ,

i 6.1.1-3, 6.1.1 -4 6 a-i. ( o ,.i T selection and fabrication i 6-2 6.3.7-1 jt-X 6.3.4Q Periodic testing of the safety injection system '

6-3 6.4-2 6.4.1.2, 6.4.2.2 Protection against the effects of toxic substance !

releases (including TMI Ill.D.3.4) 6-4 6.4-3 6.4.1.1 Control room habitability system

- 6-5 g g -6 ,t-T T W Pump seal leakage procedure 6-6 6.5-1 6.5.4.1 Containment spray system operability

- 6-7 6.2.4-1 Table 6.2.4-1 Containment isolation details 6-8 6.8.2.2.1 IRWST screen area margin analysis 7-1 7.1.2.7 Integrated response time for protection system 7-2 7.4.1.1.8.2 Operating procedures for SCS 7-3 7.3.1.1.10 Procedures for removing ESFAS signals during plant testing ;

7-4 7.3.2.1 Procedures for ESFAS Reset 7-5 7.3.2.3.2 ESFAS setpoint analyses l 7-6 7.4.2.5.2 Cold shutdown procedures l 7-7 7.5.2.5.10 Administrative controls associated with PAMI

, 8-1 8.3.1-1 8.1.4.5 Electrical power systems testing, calibration and f maintenance 3 9-1 9.1.2.2.2 Administrative controls and procedures 9.1.4.2 1. "J.hAassociated with fuel storage and handling 9.1.4.4,9.1.4.6 systems 9-2 9.2.1-1 9.2.1.1.4, 9.2.1.4, Organic fouling and inorganic buildup in the

~

9.2.5.4, 20.2.13 SSWS (including GSI-51) 9-3 9.2.1-2 9.2.1.2.1.2 Station service water system pump structure 9-4 9.2.4-1 9.2.4.2 Potable and sanitary water systems 9-5 9.3.4.1.4 Structures housing boric acid storage tank, reactor makeup water tank, and holdup tank l

Approved Design Material-Introduction Page 1.10 2 l

hu m& gest y s Y 44W k i

Sy tem 80+

.sjs ajJ nt+ 4 cca> .5-f'%.A.%e&a D sign CrntrolDocument

.4E,~<tnL M lM.

Table 1.10-1 COL License Information (Cont'd.)

COL Nc. F5ER No. Section Subject 9-6 9.5.1.5-1 9.5.1.11 Administrative controls for BTP CMEB 9.5-1

~

19.15.3.2 conformance and fire brigade

--- 9-7 9.5.1.5-1 9.5.1.12 Fire Hazards Analysis y 9-8 9.5.2-1, 9.5.2-2, 9.5.2.1, 9.5.2.2.5, Communications systems 9.5.2-3 9.5.2.2.6 j.3.u.9. i 3

-- 9-9 9.5.3.2.2 f(curity lighting system

- 9-10 9.5.4.1-1 8.3.1.1.4.11, # Diesel operator training 9-11 9.5.4.1 -2, 9.5.4.2-1, 9.5.4.2.1, 9.5.5.2, Diesel generator auxiliary support systems 9.5.5-1, 9.5.5-2, 9.5.4,j9.5.6.5, ^oj 5.c

- 9.5.6-1, 9.5.6-2, 9.5.6.2.'I, 9.5.7.2, 9.5.7-1, 9.5.8-1, 9.5.8.2.1, 9.5.9.2, 9.5.9-1 9.5.9.4, 9.5.9.5

('- 12 9.2.5-1 9.2.5.1.3 Protected area perimeter abutting or crossing a N body of water m 9-13 9.5.1.2.1.2-1 9.5.1.2 Procedures and training for using transfer switches A

VW 10-1 10.2-1 10.2.1 Turbine valve closing time 10-2 10.3-1 10.3.2.2 Steam hammer prevention 10-3 10.4.4-1 10.4.4.2.4.1 Pressure drops between the steam generator l nozzles and each system valve l 10-4 10.4.7-2, 10.4.9 2 10.4.7.2.5, Avoidance of water hammer in the condensate, fg,4 c).3,} 10.4.9.1.2 feedwater, and emergency feedwater systems 10-5 10.4.9-3 10.4.9.5.2 Steam binding in the emergency feedwater pump 11-1 11.1-1, 11.5-1 11.1, 11.5.1.1 Conformance with Appendix B to 10 CFR 20, Appendix 1 to 10 CFR 50, ANSI N13.1, R.G.

1.21 and R.G. 4.15 i 11-2 11.4-1 11.4.1.1 Site-specific solid waste management system operating procedures 11-3 11.5-2 11.5.1.4 Procedures in accordance with Position C of l R.G. 4.15

( l1-4 11.2.1-1 11.2.5 Setpoints for radiation monitors 11-5 11.5.1-1 11.5.2.6 Operation and maintenance manual for monitoring and sampling liquid and gaseous p 1[ process and effluent streams 12-1 12.1.1-1 12.1.1.2, f Operational ALARA policy 12-2 12.1.2-1 12.1.2.1, 12.2.2.1, Shielding analysis (including TMI II.B.2) 12.2.3, 12.3.2.2 12-3 12.1.3 1 h New Regulatory Guides i L

\

l- "No APeu-%

Approved Design Meterial-Introduction tf bA1 .oechin Page 1.10-3

l. - l

(-..--- - - . . -._ .-. .- .- - .. - - - - - - - . - - -

hn4 m = '.4 o %w hp ,

I Svatim 80+ Design ControlDocumart

! Table 1.10-1 COL License Information (Cont'd.)

i t

I i

COL No. FSER No. Section Subject

{

! 12-4 12.3.1-1 12.3.1.8 List of vital areas j - - - 12-5 12.3.4-1 12.3.4 Area radiation monitor location j _

12-6 12.3.4-2 12.3.4 Portable airborne iodine monitor (TMI j

III.D.3.3)

'-'"~

12 7 12.4.5-1 12.4.5 Dose assessment l

~

! 12-8 12.5-1 12.5 Health physics program 13-1 13.1-1 13.1 Organizational structure of the site operator

. 13-2 13.3-1 13.3.2 Site-specific emergency planning 13-3 13.3.3.1.7 TSC communication equipment i 13-4 13.3-2 13.3.3.2, 13.3.3.4, TMl lli.A.I.2, " Upgrade Licensee Emergency j 13.3.3.6 Support Facilities"

} 13-5 13.4-1 13.4 Review and audit I j 13-6 13.5-1 13.5 Plant procedures (TMI 1.C.1, I.C.5, & l.C.9)

I j 13-7 1.2.13, 13.6 Site security and sabotage protection j 13-8 11 2-1 13.2 Training [TMI I.A.4.2, I.A.4.1(2), ll.K.l(26)]

13-9 1s.5-2 13.5.2 Administrative Control Procedures TMI II.J.4.1, j II.K.l(10) j 14-1 14.2.2.1 Organization and Staffing i 14-2 14.2.3-1, 14.2.3-2, 14.2.1.1, 14.2.3, Testing program procedures and schedules

! 14.2.3-3, 14.2.3-4, 14.2.4, 14.2.10, including scoping documents, startup

! 14.2.4-1, 14.2.10-1, 14.2.11 administrative manual, test conditions, test j 14.2.11-1 methodologies, data collection and reduction, j reconciliation methods, and initial fuel load and j criticality procedures 14-3 14.2.6 1 14.2.6 ..etention of Test Records f

14-4 14.2.9-1 14.2.9 Trial use of plant operating and emergency

}

j procedures f 14-5 14.2.12.2-1 14.2.7.1.3 Testing of personnel monitors and radiation j survey instruments I

i 14-6 14.2.13-1 14.2.7.5 Security system detailed test description and acceptance criteria 15-1 15.7.3.4 Liquid tank failure minimum dilution flow

-l 4

15-2 15.A.3.2-1 6.1.2.2 Quantity of electrical cable insulation 17-1 17.1-1, 17.2-1 17.1, 17.2 Construction and Operation QA (including TMI j I.F.2, II.J.3.1)

{ 17-2 17.3.1-1, 17.3.5-1 17.3.1, 17.3.5, D-RAP completion

! 17.3.13 a

1 Aprowd Design Afsterial- hstroducdon Page 1.104

Syotem 80+

n4 - # b -X.4 .hM /d@

oestan controlDocument l

i 2

j Table 1.10-1 COL License Information (Cont'd.)

i e

COL No. FSER No. Section Subject f

) 17-3 17.3.9-1, 17.7 17.3.7, 17.3.9, Operations reliability assurance process i 11.3.10, 17.3.13 implementation 18-1 18.9.3.2 Validation of operating ensemble 4

j 18-2 18.6.1.3.4-1 13.2 Operator training on " Plant Safety Parameter j- Display Console" l 19-1 19.1, 19.1.2.2.2-1 19.15.3.1 Vulnerability of the intake structure due to tornado-generated debris 19-2 19.1.2.2.3-1 19.7.5.3 Elements of the plant affecting the performance of systems in seismic events j

19-3 19.1.2.4-4, 19.5, 19.15 Details of the layout of the critical components j 19.6, 19.7 for fire and flood, interaction of internal flood j sources, and effects of fire suppression systems on other systems 19-4 19.8 19.7.5.3 Development of detailed seismic walkdown

]

" Table 19.15-1 procedures to verify as-built SSC HCLPFs j 19-5 19-10, 19-11 6.5.5, 19.15 Calculation of specific flow rate and i - consideration of shielding requirements for local operator actions for the emergency containment

]

spray backup system j 19-6 19.1.2.2.6-1, 19.7.5.3, 19.15, Update of PRA to include final design detail and i 19.1.4-1, 19.1.2.4-1, 20.2.105 site-specific information including examination

, 19.1.2.4-2, of all external event hazards and analysis using 19.1.2.4-3, 19.12 site-specific spectra

]

i 19-7 19.14 19.15.6 List of risk significant SSCs for D-RAP and Table 19.15-1 operations reliability assurance process i 19-8 19.15.6 Consideration of risk imponant operator actions 19.15, Ip6,19.19 j f in developing procedures, training and human g

reliability related programs, and systems to address in severe accident management and g*

i OU .

aligning the alternate AC source (AAC) j E procedures 1

19-9 19.3.6-1, 19.3.8-1, 19.8.1.2,19.8A Establishment of administrative controls, outage I 19.3.8-2, 19.17, management, procedures and training for control l 19.18 of fire and flood barriers, contamment closure

capability and alternate equipment during shutdown operations 19-10 19.9 19.7.5 I Use of seismically rugged electrical equipment 4

4 i

1 l

) '

i

! i Approwd Design Afetenle!- Antroducelon Page 1.10-5 l

I

Sy=t:m 80 + De*1gn CrntrolDocument 2.3.4 Short-term (Accident) Diffusion Estimates (x/Q) l Atmospheric relative concentration of radiological releases is expressed as x/Q, where x is the

concentration in curies per cubic meter at the receptor and Q is the rate of release in curies per second.

4 Calculation of site-specific values of x/Q will be provided with a site-specific application that includes

, the meteorological measurements program. In lieu of site meteorological data, an assumed set of 1 4

atmospheric conditions is employed as follows to determine the values of x/Q for the System 80+ l Standard Design accident analyses.

1 i Ground-level 0 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months atmospheric dilution factors (x/Q) were calculated at 0.5 mile Exclusion Area l 1

Boundary (EAB) using a ground-level bivariate normal, or Gaussian diffusion model modified for source

configuration (i.e., building wake) and lateral plume meander under neutral and/or stable atmospheric i conditions. The methodology used in the development of accident EAB x/Q's followed the guidance a presented in NRC Regulatory Guide 1.145. Input parameters used included meteorological data representative of an 80-90th percentile U.S. commercial nuclear power plant site. For time periods greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (i.e., 0-8 hours, 8-24 hours,1-4 days, and 4-30 days), x/Q values were determined for a 2.0 mile Low Population Zone (LPZ) using logarithmic interpolation techniques which are also ,

described in Regulatory Guide 1.145. Table 2.3-1 presents the accident ground-level x/Q values at both l the EAB and LPZ receptors.

l The values of x/Q in Table 2.3-1, when combined with plant design and operational limit characteristics, !

yield acceptable doses following postulated accidents. This analytical combination is discussed in Section i 15.0.4. Details of the dose calculation methodology are given in Appendix 15A. 1 l . 2. 1 2.3.5 Long-term (Routine) Diffusion Estimates (x/Q) ;

. \

! Annual average atmospheric dilutionelative factors (x/Q) deposition and factors r&at theAvorst case (D/Q) 1 locations for various respective pathways, were calculated using the methodology as presented in NRC i Regulatory Guide 1.111 and stack release point'characteristicis provided in Table 2.3-7/ The annual I average x/Q value utilized for the residencedvhich result ' in the highest offsite dose /via the plume j mersion and inhalation pathways was 7 x 105 sec/ with a proportional D/Q value of 1.3 x 104 i . The annual average x/Q value utili d for the worst case food pathway, includin egetable, meat, l

! and milk receptors was 1.5 x 105 sec/ with a proportional D/Q value of 2.3 x 108 A site-specific j j application will determine site-specific values of long term x/Q for comparison to the values used for the System 80+ standard design routine doses. j i

, 2.3.6 Onsite (Accident) Diffusion Estimates (x/Q)

) Onsite accident I hour (i.e., applicable for 0-8 hours) atmospheric dilution factors (x/Q) were calculated

) at the north south fresh air intakes to the control room using a time based building wake model as i described by Reference 1. The north and south air intakes are located on the northwest and southeast corners of the nuclear annex, respectively. In addition, x/Qs were also calculated at the center of the

] control room and at the doorway leading to the control room itself. The x/Q values were calculated using the same meteorological database as that referenced in Section 2.3.4 along with specific source to receptor distances and building configurations used for wake considerations. For all distances, a form of the x/Q wake model associated with normal atmospheric diffusion was utilized. For time periods greater than

, 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (i.e., 8-24 hours,1-4 days, and 4-30 days) the 0-8 hour time based building wake x/Q values were adjusted using appropriate wind and occupancy factors from Reference 2. A second adjustment was made to the x/Q values so that the control room doses presented in Table 6.4-1 (calculated as described in Section 6.4.3) would represent limiting values. This adjustment consists of factors necessary to raise Approved Design Material- Site Characteristics Page 2.32

}

t E

i System 80+ Deslan ControlDocument i

{ The turbine generator placement and orientation for the System 80+ Standard Design, and the

),

corresponding low-trajectory missile strike zones, are illustrated in Figure 1.2-1. The placement and

{ orientation of the turbine generator provides adequate protection against low trajectory turbine missiles j by excluding safety-related structures, systems, and components from the low trajectory turbine missile

strike zones in accordance with the guidelines of Regulatory Guide 1.115. -

i Critical structures (i.e., those housing safety-related equipment) and exterior equipment are located in line j with, or within close proximity to, the longitudinal axis of the turbines. This makes the potential for j turbine-generated missiles to strike these targets negligibly small.

I

The System 80+ design follows the guidelines of Regulatory Guide 1.115 by placing and orienting the J turbine such that all safety-related structures, systems, and components are excluded from the low j trajectory turbine missile strike zon r if site characteristics make this impossible, safety-related targets will be placed and shielded suc that thhcombined strike and damage probability for the safety-related targets in these zones is less t r turbine failure.

I Site-specific evaluations will verify that the turbine maintenance and inspection program will ensure tiat j

the failure and missile generation probability will be less than 1.0E-4 events per turbine-year. A j summary of the turbine maintenance and inspection program and the results of the probabilistic evalualion j will be submitted for review.

j 3.5.1.4 Missiles Generated by Natural Phenomena j Tornado-generated missiles are the limiting natural hazard and, as such, are a part of the design basis for Seismic Category I structures and components. Tornado-generated missiles considered in the design are j given in Table 3.5-2.

a j 3.5.1.5 Missiles Generated by Events Near the Site -

1 l Justification will be provided in the site-rpecific SAR.

i i 3.5.1.6 Aircraft Hazards j Justification will be provided in the site-specific SAR. Also refer to Section 2.2.1.

j 3.5.2 Structures, Systems, and Components to be Protected from Externally Generated Missiles l

Tornado missiles are the design basis missiles from external sources. All safety related systems, l equipment and components required to safely shut the reactor down and maintain it in a safe condition j are housed in Seismic Category I structures designed as tornado resistant (see Section 3.5.1.4) and as j such are considered to be adequately protected.

i j 3.5.3 Barrier Design Procedures Missile barriers, whether steel or concrete, are designed with sufficient strength and thickness to stop l postulated missiles and to prevent overall damage to Seismic Category I structures. The procedures by

] which structures and barriers are designed to perform this function are presented in this section.

4 1

i i

Approvost Design Meterial- Desigrs of SSC Page 3.5-4 g.,., -

c - - - - iw--*, --- - - - - . - - - -e w.. ..,--. ._,-- - , _ _ - . , , ,,.,-s, --.w, en'r ,- , u,-,-r-

i i

System 80+ Design ControlDocument l

t 4

. e Turbine Missiles, described in Section 3.5.1.3.

I j e Natural Phenomena (Tornado) Missiles, described in Section 3.5.1.4.

! e Site Proximity Missiles (Except Aircraft), described in Section 3.5.1.5.

i e Aircraft Hazards, described in Section 3.5.1.6.

i 3.5.1.1 Internally Generated Missiles (Outside Containment) l l

Internally generated missiles (outside containment) from rotating and pressurized components are not i considered credible for the reasons discussed below. .

1

The redundant safety systems outside of containment are physically separated such that no single j gravitational or other type missile can impact both systems.

i

! 3.5.1.1.1 AmriHary Pumps and Motors i

i There are no postulated missiles originating from auxiliary pumps and associated motors outside

j. . containment for the following reasons: !

a e The pump motors are induction type which have relatively slow running speeds and are not prone

to overspeed. The motors are all pretested at full running speed by the motor vendor prior to j installation.

i e

~

j In addition to the low likelihood of missiles due to motor overspeedis discussed in A. abovebh the motor stator would tend to serve as a natural container of rotor Enssilesjf Ca =: te b: q 2 ?- % ay

!

All pumps normally have relatively low suction pressures and, therefore, would not tend to be i driven to overspeed due to a pipe break in the discharge line. In addition, the induction motor

! would tend to act as a brake to prevent pump overspeed.

I.

l e Industry pump designs are such (and service history records confirm) that there have been no I, occurrences of impeller pieces penetrating pump casings.

]

3.5.1.1.2 Emergency Feedwater Pump Turbines

}

b

j. There are no postulated missiles from the Emergency Feedwater (EFW) pump turbines for the following i reasons:

5, j e Turbine overspeed protection; electrical trip at 115 % of rated speed, and mechanical trip at 125 %

of rated speed.

i j e Assurance of turbine disk integrity by design and inspection.

i j e Enclosure of the EFW pumps and turbine drivers in a reinforced concrete room, t

i j

1 Approved Design Meteniel Desigre of SSC Pege 3.5-2 1

l' i

_ ~

CESSAREnacmu are determined by analysis and a leak detection system, as described in Section 3.6.3. l D. Through-Wall Crack l

A through-wall crack is assumed to be a circular orifice through the pipe wall of cross-sectional flow area equal to the product of one-half the pipe inside diameter and one-half the pipe wall thickness.

3.6.2.1.3 Piping Approved for Leak-Before-Break l l A leak-before-break evaluation is performed for the reactor l

coolant system (RCS) main loop piping, surge line, shutdown cooling and safety injection lines and for the main steam line inside containment which climinates the dynamic effects of pipe break from the design basis. The evaluation meets the

! requirements of 10 CFR 50, Appendix A, General Design Criterion l (GDC) 4. The evaluation is performed using the guidelines of

! NUREG-1061, Vol. 3 (Reference 1) as described in Section 3.6.3.

3.6.2.1.4 Piping Other than Piping Approved for Leak- ;

Before-Break l This section applies to all high- and moderate-energy piping

other than that whose dynamic effects due to pipe breaks are i eliminated from the design basis by leak-before-break evaluation, as identified in Section 3.6.2.1.3.

3.6.2.1.4.1 Postulated Rupture Locations A. Class 1 Piping >' l 4 _____ E3 -

(Rupture d as specified in Item p elow,[are postulated to occur at the following locations in each piping network designed in accordance with the rules of the ASME Boiler and Pressur,e Vessel Code, Section III/*(Reference 2)[for Class 1 piping:

1. The terminal ends of the pressurized portions of the run.

2. At intermediate locations selected by either one of the l following methods- -

l l

a. At each location of potential high stress and fatigue such as pipe fittings (elbows, tees, reducers, etc.), valves, flanges, and welded attachments, or

b. At each location where either of the following conditions is exceeded.

Amendment R 3.6-15 July 30, 1993 I

i i CESSAR E!%nc=,.

- Where the maximum stress range between any two load sets (including the zero load set) calculated by Eq. (10) in Paragraph NB-3653, ASME Code,Section III, exceeds 2.4 S g and the stress range calculated by either Eq. (12) or Eq. (13) in Paragraph NB-3653 exceeds 2.4 S m.

l - Where the cumulative usage f actor (U) exceeds 0.1. ,]#

Where, as defined in Subarticle NB-3650.

Sm = allowable stress-intensity value.

U = the cumulative usags factor.

As a result of piping reanalysis due to differences between

, the design configuration and the as-built configuration, the l highest stress or cumulative usage factor locations may be I

shifted; however, the initially determined intermediate break locations need not be changed unless one of the following conditions exists:

(i) The dynamic effects from the new (as-built) intermediate break locations are not mitigated by the original pipe whip restraints and jet shields.

l (ii) A change is required in pipe parameters such as major

! differences in pipe size, wall thickness, and routing.

Leakage crack locations for Class 1 piping are specified in l Item F below.

B. Class 2, Class 3, or Seismically Analyzed ANSI B31.1 Piping

[ Ruptures),asspecified Item below, bre postulated to occur at the following rocati s in each piping network designed PressureinVessel accordance Code,with the rules Section of the ASME2)[ Boiler II [ (Reference for Class and 2 and Class 3 piping, or with the rules of the ASME Code for

/

I h

b For those loads and conditions in which Level A and Level B stress limits have been specified in s the design specification (excluding earthquake loads). f Amendment S 3.6-16 September 30, 1993

=

l i

CESSAR MEnema l

Pressure Piping, B31, Power Piping, ANSI /ASME B31.12 (Reference 3)(for seismically analyzed ANSI B31.1 piping:

1. the terminal ends of the pressurized portion of the j network, and

2. either

/

a. intermediate locations of potential high stress or i fatigue such as pipe fittings, valves, flanges and

! welded-on attachments, or l

b. where the piping contains no fittings, weld

. attachments, or valves, at one location at each extreme of the piping run adjacent to the i protective structure, or l

c. intermediate locations where the stress, S, exceeds 0.3(X+Y).JP where, as defined in Subarticle NC-3650, S = stresses under the combination of loadings for which either Level A or Level B service limits have been

[

specified, as calculated

from the sum l of equations (9) and (10).

X = equation (9) Service Level B allowable stress.

l Y = equation (10) allowable stress.

As a result of piping reanalysis due to differences between the design configuration and the as-built configuration, the l highest stress locations may be shifted; however, the initially determined intermediate break locations may be used unless a redesign of the piping resulting in a change in pipe parameters (diameter, wall thickness, routing) is required, or the dynamic effects from the new (as-built) intermediate break locations are not mitigated by the original pipe whip restraints and jet shields.

Leakage crack locations for Class 2 and Class 3 piping are specified in Item F below.

l l

i For those loads and conditions in which Level A and Level B l stress limits have been specified in the design l

specification (excluding earthquake loads).

i Amendment S I

3.6-17 September 30, 1993

- =

l r

CESSAR !!!nncmou !

1 l

C. Non-Safety Related ANSI B31.1 Piping System 80+ piping is designed so as to isolate seismically analyzed piping from non-seismically analyzed piping. In cases where it is not possible or practical to isolate the seismic piping, adjacent non-seismic piping is analyzed according to Seismic Category II criteria. For non-seismic piping attached to seismic piping, the dynamic effects of the non-seismic piping are simulated in f.he modeling of the i seismic piping. The attached non-seismic piping up to the l analyzed /unanalyzed boundary is designed not to cause a !

failure of the seismic piping during a seismic event. j For non-safety class piping which is not seismically analyzed, leakage cracks are postulated at axial locations such that they produce the most severe environmental effects.

D.[ Break Locations in Piping Runs with Multiple ASME Code Piping Classes

Breaks [ in accordance with Section 3.6.2.1.4.1.E,[are j postulated to occur at the following locations

l

1. The terminal ends of the pressurized portions of the l

run. 7 V

2. At intermediate locations selected by either one of the following methods:

i

a. At each location of potential high stress or fatigue, such as pipe fittings, valves, flanges, and welded attachments; or

b. At all intermediate locations between terminal ends where the stress and fatigue limits of l l Sections 3.6.2.1.4.1.A.2.b or 3.6.2.1.4.1.B.2.c i

areexceeded.f Break Locations E.

Both circumferential and longitudinal breaks are postulated to occur, but not concurrently, in all high-energy piping systems at the locations specified in Items A, B, C, or D, except as follows:

l 1. Circumferential breaks are not postulated in piping l runs of a nominal diameter equal to or less than 1

inch.

2. Longitudinal breaks are not postulated in piping runs l of a nominal diameter less than 4 inches.

Amendment V

' <-10 April 29, 1994

CESSAREnLb-l 4

3. Longitudinal breaks are not postulated at terminal ends.

4. Only one type of break is postulated at locations where, from a detailed stress analysis, such as finite-element analysis, the state of stress can be used to identify the most probable type. If the primary plus secondary stress in the axial direction is found to be l at least 1.5 times that in the circumferential i direction for the most severe loading combination i association with Level A and Level B service limits, then only a circumferential break is postulated.

Conversely, if the primary plus secondary stress in the circumferential direction is found to be at least 1.5 l times that in the axial direction for the most severe loading combination associated with Level A and Level B ser vice limits, then only a longitudinal break is postulated.

1

5. Circumferential and longitudinal breaks are not postulated at locations where the requirements of Item G are satisfied. l

6. Circumferential and longitudinal breaks are not l postulated at locations where the criterion in Item F.2 is used, l

F. Crack Locations l I

1. Through-Wall Cracks

[Through-wallcracksarepostulatedinallhigh-energy ;

and moderate-energy piping systems ,having a nominal

/ l diameter greager than 1 inch i

^

m.: cpec; lied 1 1. A- " w h except that through-wall cracks are not postulated at locations where:

a. For Class 1 piping, the calculated

value of S, as defined in Item A, is less than one-half the limits of Item A.2.b.

b. For Class 2, Class 3 or seismically analyzed ANSI B31.1 piping, the calculated

values of S as defined in Item B.2.c is less than one-half the limits of Item B.2.c. [

c. The requirements of Item G are satisfied.

d. The criterion in 2. below is used.

For those loads and conditions in which Level A and Level B

. stress limits have been specified in the design specification (excluding earthquake loads).

Amendment V 3.6-19 April 29, 1994

=

1 -

CESSAR HiMemou circumferential breaks are postulated in fluid system piping and branch runs as specified in Section 3.6.2.1.4.1.E.

Instrument lines, one inch and less nominal pipe of tubing size are designed to meet the provisions of Regulatory Guide 1.11.

Longitudinal breaks in fluid system piping and branch runs are postulated as specified in Section 3.6.2.1.4.1.E.

B. Crack Configurations l Through-wall cracks are postulated at those axial locations specified in Section 3.6.2.1.4.1.F.

For high-energy piping, through-wall cracks are postulated to be in those circumferential locations that result in the most severe environmental consequences. The flow from the crack is assumed to wet all unprotected components within the compartment with consequent flooding in the compartment and communicating compartments. Flooding effects _ are determined on the basis of a conservatively estimated time period required to effect corrective actions.

l 3.6.2.1.5 Details of Containment Penetrations Details of containment penetrations are discussed in Sections 3.8.1 and 3.8.2.

3.6.2.2 Analytical Methods to Define Forcine Functions and Response Models 3.6.2.2.1 Piping Approved for Leak-Before-Break There are no forcing functions or response models for the reactor coolant loop, surge line, shutdown cooling line, safety injection l

line and main steam line based upon elimination of dynamic effects by leak-before-break evaluation.

l 3.6.2.2.2 Analytical Methods to Define Forcing Functions I and Response Models for Piping Excluding That j

Approved for Leak-Before-Break l

This section applies to all high-energy piping other than that whose dynamic effects due to pipe breaks are eliminated from the design basis by leak-before-break evaluation.

3.6.2.2.2.1 Determination of Pipe Thrust and Jet Loads l

A. Circumferential Breaks ,

/

P(Circumferential breaks are assumed to result in pipe least a severance and separation amounting to at Amendment S 7.6-74 Sectember 30, 1993

CESSARinecum one-diameter lateral displacement of the ruptured piping sections, unless physically limited by iging restraints, structural members, or piping stiffness.

2 1 B. Dynamic Force of the Fluid Jet Discharge l CThe dynamic force of the fluid jet discharge from either a I postulated circumferential or longitudinal break is based on a circular break area equal to the cross-sectional flow area of the pipe at the break location and on a calculated fluid /

4 pressure modified by an analytically determined thrust coefficient, as determined for a circumferential break at the same location. D ine restrictions, flow limiters, positive pump-controlled flow, and the absence of energy reservoirs are taken into account, as applicable, in the reduction of jet discharge.

Piping movement is assumed to occur in the direction of the ,

jet reaction, unless limited by structural members, piping l I

restraints, or piping stiffness.

a v. Pipe Blowdown Force and Wave Force

[Thefluidthrustforcesthatresult from either postulated /

circumferential or longitudinal breaks, are calculated psing a simplified one step forcing function methodcMgy.] This methodology is based on the simplified methods described in References 5 and 6. I the simplified method]d' discussed above (leads to 1

{When impractical protective measures, then a more detailed r )

computer solution which more accurately , reflects the j postulated pipe rupture event is used.] The computer '

solution is based on the NRC's computer program developed

. for calculating two-phase blowdown forces (Reference 7).

D. Evaluation of Jet Impingement Effects 1

Jet impingement force calculations are performed only if structures or components are located near postulated high ,

energy line breaks and it cannot be demonstrated that failure of the structure or component will not adversely affect safe shutdown capability.

E. Longitudinal Breaks A longitudinal break results in an axial split without severance. The split is assumed to be orientated at any point about the circumference of the pipe, or alternatively at the point of highest stress as justified by detailed /

stress analyses. For the purpose of design, [the longitudinal break is assumed to be circular or elliptical (2D x 1/2D) in shape, with an area equal to the largest

~

Amendment S 3.6-25 September 30, 1993

l l

CESSAR EnWicmox ;

1 piping cross-sectional flow area at the g nt of the break and have a discharge coefficient of 1.0.J'Any other values l l used for the area, shape and discharge coefficient associated with a longitudinal break is verified by test data which defines the limiting break geometry.

3.6.2.2.2.2 Methods for the Dynamic Analysis of Pipe Whip Pipe whip restraints usually provide clearance for thermal expanLion during normal operation. If a break occurs, the restraints or anchors nearest the break are designed to prevent unlimited movement at the point of break (pipe whip) The dynamic nature of the piping thrust load is considered. b n the absence of analytical justification, a dynamic load factor of 2.0 /

is applied in determining restraint loading 7 Elastic-plastic pipe and whip restraint material properties may be considered as applicable. The effect of rapid strain rate of material

! properties is considered in accordance with Reference 5. A 10 percent increase in yield strength is used to account for strain rate effects.

3 In general, the loading that may result from a break in piping is

- determined using either a dynamic blowdown or a conservative static blowdown analysis. The method for analyzing the interaction effects of a whippi{ng pipe with a restraint is one of ,/

the following: (1) the Energy Balance Method (2) Lumped Parameter Method, or (3) Equivalent Static Method.fe-The energy balance method is based on the principle of conservation of energy. The kinetic energy of the pipe generated during the first quarter cycle of movement is assumed to be i converted into equivalent strain energy, which is distributed to l the pipe or the whip restraint. See Appendix 3.6A for a discussion of the application of the energy balance method. <

i I f

The lumped parameter method is carried out by utilizing a lumped 4 mass model. Lumped mass points are interconnected by springs to

take into account inertia and stiffness properties of the system.

A dynamic forcing function or equivalent static loads may be i

applied at each postulated break location with pipe whip interactions. A nonlinear elastic-plastic analysis of the piping-restraint system is used. The computer method for this analysis is described in Appendix 3.6A.

A conservative static analysis model is used for rigid rupture restraints. In order to obtain the design load for a rigid restraint, the following equation is used:

F = 2 x 1.1 x F,

= 2.2F, 4

Amendment U 3.6-2A December 31, 1997 )

~

CESSAR 8nG"icmt.

J )

where F = the design load F, = maximum blowdown force and the dynamic load factor (DLF) is taken as 2.0 and rebound effects are accounted for by a factor of 1.1.

3.6.2.2.2.3 Method of Dynamic Analysis of Unrestricted Pipes

,/

[ The impact velocity and kinetic energy of unrestricted pipes is calculated on the basis of the assumption that the segments at each side of the break act as rigid-plastic cantilever beams The hinge subject to piecewise constant blowdown forces.

location is fixed either at the nearest restraint or at a point ,

)

determined by the requirement that the shear at an interior plastic hinge is zero.Khe kinetic energy of an accelerating cantilever segment is equal to the difference between the work done by the blowdown force and that done on the plastic hinge.

The impact velocity V. is found from the expression for the kinetic enensv:

KE = (1/2) M ,V '2 where M, is the mass of the single degree of freedom dynamic model of the cantilever. The impacting mass is assumed equal to M., .

For a straight run of pipe rotating about a plastic hinge, the zone of influence of the whipping pipe accounts for an increasing length due to a traveling hinge point caused by strain hardening effects. The impact energy of unrestrained pipe ir,to a barrier (e.g. the divisional wall) is governed by the vector component of its velocity at impact which is perpendicular to the barrier. l Impact of small piping into building structures conservatively l' assumes that all of the impact energy is imparted to the barrier with no dissipation due to local crushing deformation of the l

pipe. l Bearing area of impact on building structure is generally elliptical, but is treated as a circle of equivalent area, with i dimensions based on experimental data for pipe crush behavior. !

As the impact load is greatest on the periphery of the ellipse, this yields a conservative force distribution into the barrier.

Long term loading on the barrier subsequent to impact due to system blowdown and continued deceleration of remaining pipe (beyond the impact zone) is accounted for in addition to the initial impulsive loading. -

Amendment P 3.6-27 June 15, 1993

CESSAR SPecma i

i The change in J-integral with crack length "a" is determined by analyzing several crack lengths in the region of interest. For a leakage crack of length "a", crack lengths "a", a-6, and a+6 ,

are analyzed. Similarly, the change in J-integral with crack ;

length in the region of length "2a" is determined by analyzing i cracks with lengths 2a, 2a-6, and 2a+6. This method provides the derivative information in the two regions of interest. The variation of J with crack length in the region of "a" and "2a" is plotted along with the material curve. Evaluation of the plots allows for direct verification of the stability criteria.

l The evaluations are performed for the locations chosen to envelop all limiti'ng cases. The pipes with the leakage crack length 4 subject to loads of 4 x (P+NOP+M ::ir.um Design Load) and the pipes with crack length twice the leakage crack length with loads of (P+NOP+ Maximum Design Load) are demonstrated to have l significant margin between the material curve and the loading curve, indicating that all pipe locations satisfy the LBB crack stability criteria.

See Appendix 3.9A, Sections 1.1.9.5.4 and 1.1.9.6 for a discussion of LBB design criteria development and a further discussion of analytical methods.

3.6.3.8 Results , _

The piping listed in Section 3.6.3 and evaluated by the methods described above are shown to meet all the [ criteria for /

application of eak-before-breakJagcording to NUREG 1061, Volume 3. Speci Ically, these criteria (require that:

l 4.[Crackswhichareassumedtogrowthroughthepipewallleak significantly while remaining stable. The amount of[ leakage /

is detectable with a safety margin of at least a factor of 10[nlessotherwisejustified.

B. Cracks of the length that leak at the rate in A. can withstand normal operation plus maximum design load loads with a safety factor of at least 4 I

c. Cracks twice as long as those addressed in B. will remain /

stable when sub to normal operation plus maximum designload.]*T*jected Site specific evaluations will confirm that the bases for the LBB l acceptance criteria are satisfied by the final as-built design and materials of the piping systems listed in Section 3.6.3 and will be documented in a LBB evaluation report.

Amendment W 3.6-43 June 17, 1994

CESSAR Once l l

j include motions due to rocking and torsion. The total seismic response of the piping is then calculated by absolute summing the !

i results of the response spectrum analysis and a static analysis i 4

which accounts for the relative displacement effects between l support locations. Since the displacement effects are self- :

limiting, it is justified to place them in the secondary stress I category.

As an alternative to the modal response method, a time history !

! method of analysis may be used. This method is also used for i other types of dynamic analyses such as LOCA and hydraulic 1 transients. Either a direct integration method or a modal j superposition method is used to solve the equations of motion.

3.7.3.2 Determination of Number of Earthauake Cycles The procedure used to account for the fatigue effect of cyclic i motion associated with seismic excitation recognizes that the actual motion experienced during a seismic event consists of a

, single maximum or peak motion, and some number of cycles of lesser magnitude. The total or cumulative usage factor can also be specified in terms of a finite number of cygles of the maximum or peak motion. Based on this consideration,/. Seismic Category I subsystems, components, and equipment are designed for a total of two SSE events with 10 maximum stress cycles per event (20 full #

cycles of the maximum SSE stress range). Alternatively, an 4 equivalent number of fractional vibratory cycles to that of 20 !

1 full SSE vibratory cycles may be used (but with an amplitude not )

less than one-third (1/3) of the maximum SSE amplitude) when ,

derived in accordance with Appendix D of IEEE Standard 344-1987.

3.7.3.3 Procedure Used for Modelinc The modeling techniques incorpora?.e either a single or multidegree of freedom subsystem consisting of discrete masses j connected by spring elements. The associated damping i

coefficients are consistent with Table 3.7-1. The degree of complexity of each model is sufficient to accurately evaluate the i dynamic behavior of the component. For additional details on pipe modeling, see the section below.

Valves (i.e. , with natural frequencies greater than the frequency

. corresponding to the zero period acceleration (ZPA)) are included in the piping system model as lumped masses on rigid extended structures. If it is shown by test or analysis that a valve has a frequency less than a frequency corresponding to the ZPA, then a multimass, dynamic model of the valve, including the appropriate stiffnesses, is developed for use in the piping system model.

l Amendment R 3.7-20 July 30, 1993

- .- 1 CESSAR unince,. l Torsional effects of eccentric masses are also considered in the ,

analysis of Seismic Category I subsystems other than piping. !

l 3.7.3.12 Pipinc Outside Containment Structure 3.7.3.12.1 Buried Piping Class 2 and 3 buried piping systems are designed according to the seismic analysis acceptance criteria for Seismic Category I buried piping systems. These criteria accept either dynamic analyses or equivalent static load methods with the consideration of seisr.ic effects which are induced primarily by seismic wave 1 passag.a and by differential movements between building attachment pointa and the ground surrounding the buried pipe.

Seismic effects in buried piping are self-limiting (displacement-induced strains and associated stresses) rather than being in i equilibrium with an external load. In accordance with the ASME l B&PV Code stress classification rules, the elastically calculated seismic stresses in buried piping, which are the products of ,

strains and Young's moduli of the piping materials, are secondary j stresses rather than primary stresses in nature. The design and stress acceptance criteria as provided below have been modified from the Code Subarticles NC/ND-3600. Nomenclatural symbols are defined in the Code unless noted otherwise.

3.7.3.12.1.1 General Requirements ,

General design requirements for ABB-CE System 80+ buried piping systems are as f>llows: l A. Areas of direct fault displacement and unstable soil conditions, such as liquefaction, are avoided for buried pipe installation.

B. [Conformance to allowable structural and piping stresses after the line penetrates the Nuclear Annex is assured b /

the use of either expansion joints or flexible seals. J#y C. Minimum pipe wall thickness is per Code Equation 3 of l )

NC/ND-3641.1 as follows: l PD m " 2(S + Py) *^

Amendment Q 3.7-24 June 30, 1993

l CEOSAR MEnema i

10.4.2.3 Design Loads (Reference Section 3.8.3.3) l Refer to Table 3.8A-1 for additional loads that are applicable to the Crane Wall.

10.4.3 REFUELING CAVITY 10.4.3.1 Description l

Re Refueling Cavity is the reinforced concrete enclosure that provides a pool filled with borated l l -- water above the reactor vessel to facilitate the fuel handling operation without exceeding the l l acceptable level of radiation inside the Containment Vesss he Refueling Cavity has the following {

sub-compartments.

Storage Area for Upper Guide Structure ,

Storage area for Core Support Barrel l

Refueling Canal l l

l The Reactor Vessel flange is sealed to the bottom of the Refueling Cavity to prevent leakage of refueling water into the reactor cavity. He Fuel Transfer Tube connects the Refueling Cavity to the Spent Fuel Pool. The shield walls that form the Refbeling Cavity are a minimum of six feet thick.

10.4.3.2 Design Requirements he Refueling Cavity walls and floor shall be covered with stainless steel plate for leak tightness and for contamination and corrosion control.

l 7 10.4.3.3 Design Loads (Reference Section 3.8.3.3)

ReSueIing Cus,+y .

Refer to Table 3.8A-1 for additional design loads that are applicable to the C= .; "?f ;

10.4.4 OPERATING FLOOR I 10.4.4.1 Description l

l The Operating Floor at El.146'-0" provides access for operating personnel functions and provides biological shielding. Inside the Crane Wall, the operating floor is a reinforced concrete slab with a covered hatch that is aligned with hatches in the two lower floors. Outside the Crane Wall, the Operating Floor consists of steel grating.

I 10.4.4.2 Design leads (Reference Section 3.8.3.3)

Refer to Table 3.8A-1 for additional design loads that are applicable to the Operating Floor.

l l

l Amendment V I l 3.8A-38 April 29,1994

i 1

I System 80+ &&n contrat Document Table 3.8A-6 Radwaste Facility, SSE Accelerations in Gs Elevation NS EW Vertical Roof (by COL) (by COL) (by COL)

Second Floor (by COL) (by COL) (by COL)

First Floor (by COL) (by COL) (by COL)

Basemat (by COL) (by COL) (by COL) l l

1 TaEe 3.8A-7 Dead Weight Loads for M40r Turbine Building Equipment l

Weight Location Item Quantity (Kips)

)

Ground Floor CondensersD3 3 2750 Operatmg Floor Reheater 2 450 Reheater Drain Tank 2 27

)

Ist Stager21 Reheater Drain Tank 2 27 2nd Staget21 Operating Floorfrurbine Law Pressure Turbine 3 8500 Pedestal g

Operating Floor I Deaerator FW StodTQb 1700 MS Drain Tankt21 2 31 Mezzanine Floor PT Heater 4 290 UI includes LP Feedwater Heaters l23 Suppons below the operating floor The above weights are the operating weights of each item and includes the weight of contained fluids, fT specific SAR shall verify the above information based upon the turbine purchased.)] 3 Anwend Dee> AtaterW- Design of S Pope 3.8A-66 3 COL inOneko'^ *;det lch G bbEb &cb4 M.

I

CESSAR n! Dime,.

The following loading combinations from Table 3.8-5 and Appendix 3.8A are used for analysis and design of Category I structures and their components. See Appendix 3.8A. Section 5.0 for definitions.

4.1 LOADING CO51BINATIONS FOR SEIS3flC CATEGORY I CONCRETE STRUCTURFJ 4.1.1 SERVICE LOAD COMBINATIONS

~~

a) U = 1.4D + 1.7L b) U = 1.2D + 1.7W c) U ,1.4D + 1.7F + 1.7L + 1.7H + 1.7W d) U = (.75)(1.4D + 1.7F + 1.7L + 1.7H + 1.7To + 1.7R,)

e) U = (.75)(1.4D + 1.7F + 1.7L + 1.7H + 1.7W + 1.7To 1.7R,)

4.1.2 FACTORED LOAD COMBINATIONS 4

a) U = D + F + L + H + T + R, + E' b) U = D + F + L + H + To+ R, +W, c) U = D + F + L + H + T + R, + 1.5 P, d) U = D + F + L + H + T + R, + 1.0P, + 1.0(Y,+ Y3 +Y,) + E' 4.2 LOADING COMBINATIONS FOR SEISMIC CATEGORY I STEEL STRUCTURES 4.2.1 SERVICE LOAD CONDITIONS 4.2.1.1 Elastic Allowable Strength Design

! a) S=D+F+L+H b) S=D+F+L+H+W c) 1.3 S = D + F + L + H + R, + To d) 1.3 S = D + F + L + H + W + R, + T, 4.2.1.2 Plastic Design a) Y = 1.7 (D + F + L + H) b) Y = 1.7 (D + F + L + H + W) c) Y = 1.3 (D + F + L + H + T, + R,)

d) Y = 1.3 (D + F + L + H + W + T, + R,)

4.2.2 FACTORED LOAD CONDITIONS I

4.2.2.1 Elastic Allowable Strength Design l a) 1.4 S = D + F + L + H + R, + To + E' l b) 1.4 S = D + F + L + H + Ro + To + W, c) 1.4 S = D + F + L + H + R, + T, +' P, d) 1.6 S = D + F + L + H + R, + T, + (Y,+ Y)+ Ym ) + E' + P, 3.8B-2 Amendment U - 12/31/93 l

l

l System 80+ Design ControlDocument

4. The upper lift coil is energized moving the de-energized upper latch assembly up 7/16 inches.

6 /~5Z. M i

5. The upper latch coil is energized engaging the latches with clearance.

sis cut,.

6. The lower lift coil is de-energized allowing the lower latch to drophe $.: die ahan.' w - -

The drive shaft will move downgch, stopping on the upper latch assembly, which is energized and in its up position, u/ n.

7. The lower latch coil is de-energized disengaging the lower latches.

8. The upper lift coil is de-energized lowering the upper latch assembly with the drive shaft M inch.

E *1/ K.

3.9.4.2 Applicable CEDM Design Specifications The pressure boundary components are constructed in accordance with the requirements for Class 1

, vessels per the applicable Edition and Addenda of Section III (Subsections NCA and NB) of the ASME Boiler and Pressure Vessel Code and the criteria of SRP 3.9.4, Rev. 2 Subsection 11.2. The pressure boundary material complies with the requirements of Section Ill and IX of the ASME Boiler and Pressure l Vessel Code and Code Case N-4-II.

The adequacy of the design of the non-pressure boundary components has been verified by prototype accelerated life testing as discussed in Section 3.9.4.4.

The reed switch position transmitter (RSPT) assembly of the CEDM is designed to comply with IEEE Standard 323-1974, standard for " Qualification of Class I Electrical Equipment for Nuclear Power Generating Stations," and IEEE Standard 344-1987, " Recommended Practice Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations", as endorsed by Regulatory Guide 1.100 Rev. 2, dated June 1988. The electrical components are external to the pressure boundary and are non-pressurized.

The test program to verify the CEDM design is discussed in Section 3.9.4.4.

3.9.4.3 Design Loads, Stress Limits and Allowable Deformations The CEDM stress analyses consider the following loads:

Reactor coolant pressure and temperature

Reactor operating transient conditions e Dynamic stresses produced by seismic loading and design bases pipe breaks and/or LOCA loading.

e Dynamic stresses produced by mechanical excitations

, Full length RSPT assemblies are subjected to biaxial random multi-frequency input motions corresponding j to design bases excitations. Testing is performed using four RSPT orientations to account for asymmetries in the design.

Approved Design Metenal- Design of SSC Page 3.9-47 l

CESSAR !!E"icamn l l 1

I TABLE 3.9-10 I 4

LOADING COMBINATIONS FOR ASME SECTION III CLASS 1 PIPING 8ervice Love,1 Loadinct combination Design Design Pressure, Weight, Other Sustained Mechanical Loads Level A Level A Transients, Weight, Operatincj Pressure, Thermal Expansion, Anchor Movements, other Mechanical Loads, l l Dynamic Fluid Loads l Level B Level B I Transients, Weight, coincident Pressure, Thermal Expansion, Anchor Movements, /

[ Safe Shutdown Earthquake,2'37

sother Mechanical Loads, Dynamic Fluid Loads Level C Maximum Pressure, Other Mechanical Loads, Weight, Dynamic Fluid Loads Level D Maximum Pressure, l

Other Mechanical Loads, Weight, i

Safe Shutdown Earthquake,

! Pipe Break Loads, l l i [ Dynamic SSE SAMS Fluid (FullLoads Ran Thermal TAMS,2 /

Thermal Expansion *ge)J' NOTES: The dynamic loads are combined by the square root of the sum of the squares.

[l Alternatively, a lower level of SSE motion may be used in accordance with Section J.7.3.2.

2 Loading combination for Eq. 12a of Reference 50. l [

3 Primaryplussecondarystressproducingload]Y i

j,a,m nanen & q.pD l

Amendment Q June 30, 1993

,n - ,- -

i CESSAREnnean 1 TABLE 3.9-11 j l

LOADINO COMBINATIONE FOR ASME SECTION III CLASSES 2 AND 3 PIPING 8ervice Level Loadine combination Design Design Pressure, Weight Level A & B Operating Pressure, l Weight, other occasional Loads (DFL, Wind)

Thermal Expansion, Anchor Movements Level C Maximum Pressure, l Weight, other Occasional Loads (DFL, ,

Tornado) l Level D Maximum Pressure, l Weight, DFL, Safe Shutdown Earthquake, Pipe Break,

( Anchor Movements1 ,' -

Thermal Expansionk]* l l

l NOTES: Dynamic fluid loads (DFL) are occasional j loads such as safety / relief valve thrust, ;

steam hammer, water hammer, or loads '

associated with plant upset or faulted condition as applicable.

[1 Loading Combination for Eq. 10b of Reference '

50. ]{fP i

Amendment Q June 30, 1993

$3 AS$fM81.f GutDE FA7At i

o ,- - n ., 7 GUIDE STRUCTURE l SUPPORT SYSTEM 3 / -- .

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erv Amendment U - 12/31/93 9**

UPPER GUIDE STRUCTURE ASSEMBLY Jg _

yY System 80+ Design Control Document i 1.4.3.2.2 Time IIistory Analysis 1.4.3.2.2.1 General Time history analysis is used as an alternative method to response spectrum analysis for any piping '

system.

For those piping systems analyzed by time history methods, development of mathematical models, which define flexibility and mass, and calculation of natural frequencies and mode shapes, as described in ,

Section 1.4.3.2.1.1 of this appendix, is first performed. l

)

1.4.3.2.2.2 Piping Dynamically Decoupled from the NSSS Most piping systems are dynamically decoupled from the nuclear steam supply system (NSSS), following i the guidelines of Section 1.5.2.2 of this appendix.

The solution of the differential equations of motion, which describe the dynamic response of a system ,

to a seismic excitation, is obtained by the method of modal superposition or by the method of direct i integrations, using time history analysis. These methods are described in Section 3.7.2.1.1.2. l When modal superposition is used, the number of modes analyzed is se'.:cted to account for the principal j~ vibration modes of the system based upon mass and stiffness properGes, modal participation factors and closeness to amplified region of the response spectrum of the input time history. As required on a case by case basis, the analysis is repeated with more modes in order to verify the chosen number of modes.

For time history analysis by direct integration, the integration time step is chosen to be a maximum of l 1/10 of the period of the highest frequency of the system. As required on a case by case basis, the time step is reduced and the analysis repeatrJ in order to verify the choice of time step. When time history analyses are performed for subsystems by direct integration, the time step is shifted up to il5% to account for uncertainties.

E r

-hme hNSW4Ot.freguencycoriht The mathematical model is subjected to seismic exci DTIM: :M -ciitFr points (terminal ends) and at building supports. For statistically independen: earthquake motions, input excitations in all three translation directions and, as applicable, in all three rotational directions are applied simultaneously to the anchor points and building supports. '

Input of multiple time history excitations, which allow calculation of the effects of both differential motion and inertia, are used in a multi-supported system such as a piping system. An alternate time history method used, as described in ASME Code, Section Ill, Division I, Appendix N, Sc: tion N-1228.4, is to input an " envelope" time history excitation to calculate the inertia response, and separately to determine the effects of differential support motion using a static analysis. The ASME B&PV Code defines the envelope excitation as a time history whose response spectrum envelopes the response spectra l for the individual support motions.

1.4.3.2.2.3 Piping Dynamically Coupled to the NSSS The only piping system that is dynamically coupled to the NSSS for the purpose of structural analysis is the main coolant loop piping. The main coolant loop piping is seismically analyzed as an integral part of the reactor coolant system structure, using methods described in Sections 3.7.2.1.2 and 3.7.2.6.2.

Approved Design Meteriel- Design of SSC Page 3.9A-8

[

System 80+ Design ControlDocument 1

l 1.5 Analysis Tecchniques 1.5.1 Model Boundaries Piping models ideally run from anchor to anchor (equipment nozzle, or penetration). Where this is not feasible, the piping is separated by decoupling, overlapping, isolation, or in-line anchors as described in the following subsections to form more manageable models for analysis. Where the piping cannot be separated to form smaller analysis models by these methods, the use of an intermediate anchor is considered in order to separate models, subject to the considerations of Section 1.5.5 of this appendix.

1.5.2 Decoupling 1.5.2.1 General Small branch lines are allowed to be decoupled from larger run piping regardless of seismic classification.

In some instances, decoupling is also applied for in-line pipe size changes (such as at a reducer or reducing insert). In the description in Section 1.5.2.2, the smaller line is defined as the " branch" and the larger line is defined as the "run" To meet decoupling criteria, piping meets the size or moment of inertia ratios as detailed in the following paragraphs. For decoupling criteria to be meaningful, the branch line must be designed flexible enough to absorb the anchor motions of the run pipe. Therefore, the branch line flexibility is maintained by avoiding placement of branch line supports close to the run pipe.

1.5.2.2 Branch Decoupling Criteria ilBranch lines meeting thefollowing criteria may be decoupledfrom t i.e main run:

l Dy'D, & 0.33, or 4/I,10.04,))'

where:

l D3 = Branch nominal pipe size D, = Run nominal pipe size 13 = Branch moment of inertia I, = Run moment of inertia f,

\(An appropriate stress intensityfactor (SIF) is included on the branch and main run lines af the point where thepiping is decoupled. Mass effects of the branch line are considered in the analysfs of the run h.

line. The branch point is considered as an anchor in the analysis of the branch pipejThermal and iN seismic anchor movement analyses of the decoupled branch lines areperformed with the thermal, seismic 4 . '

inenial, seismic anchor movement (SAM), orpipe break movements of the largerpipe header applied as 3

NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5.

Approved Design Material- Design of SSC page 3 gA.14 l

1 CESSARML%ma

, e/

(Anappropriat tress intensity factor (SIF) is included on the branch an main run lines at the poir.t wher 'the piping is decoupled ass effects of the branch line ar onsidered in the '

analysis f the run line. The branch poi s considered as an i anchor in the analysis of the branch pipe .hermal and seismic S

anchor movement analyses of the decoup d branch lines are performed with the thermal, seismic inertial, seismic anchor j movement (SAM), or pipe break movements of the larger pipe header y applied as anchor displacements and/or rotations to the smaller branch line wherever these movements are significant. The inertia effects of the run pipe on branch pipe are considered,

! where significant.]#-

l Piping is also decoupled at flexible hose wherever each interfacing analysis considers the flexible hose weight and significant stiffness, and wherever the flexible hose qualifies for the net end displacements of the interfacing analysis problems. Analysis results of the interfacing problems are not combined. The flexible hose is not allowed to experience loads beyond those recommended by the manufacturer.

Also refer to Section 3.7.2.3.3 for general decoupling criteria.

1.5.2.3 Seismic to Non-seismic Decoupline Criteria Two methods for designing the region of a seismic /non-seismic i piping interface are as follows:

A. Use of structural anchors for isolation i

l Structural isolation anchors provide an effective means of

! protecring seismic piping from the seismic response of non-seismically designed piping. Anchors are designed assuming that a plastic hinge forms at the interf ace with non-seismic piping.

i

! B. Use of isolation restraints Piping restraints are utilized to isolate the seismic response of non-seismically designed piping from seismically designed piping. Isolation restraints tre designed as follows:

i j . Two restraints located in each of the three orthogonal directions are used to isolate the seismically caused pipe moments, and forces from the non-seismic piping to the seismic piping.

1 Amendment T 3.9A-18 November 15, 1993

i i System 80+ Design controlDocument ,

p'o anchor facements and/or rotations to the smaller branch line wherever these movements are "

ihm sigmficant. The inertia effects of the run pipe on branch pipe are considered, where sigmficant.' l Piping is also decoupled at flexible hose wherever each interfacing analysis considers the flexible hose weight and significant stiffness, and wherever the flexible hose qualifies for the net end displacements of the interfacing analysis problems. Analysis results of the interfacing problems are not combined. The , flexible hose is not allowed to experience loads beyond those recommended by the manufacturer. l l 1 Also refer to Section 3.7.2.3.3 for general decoupling criteria. j e l 4 1.5.2.3 Seismic to Non-seismic Decoupling Criteria Two methods for designing the region of a seismic /non-seismic piping interface are as follows:

1. Use of structural anchors for isolation i i

Structural isolation anchors provide an effective means of protecting seismic piping from the seismic response of non-seismically designed piping. Anchors are designed assuming that a plastic hinge forms at the interface with non-seismic piping.

2. Use of isolation restraints Piping restraints are utilized to isolate the seismic response of non-seismically designed piping from seismically designed piping. Isolation restraints are designed as follows:

Two restraints located in each of the three orthogonal directions are used to isolate the seismically caused pipe moments. and forces from the non-seismic piping to the seismic piping.
An isolation restraint is designed for anticipated seismic loads including the additional loads resulting from the potential failure of the non-seismic piping and pipe supports.
The stress allowables given in ASME Section III NF for Level D loadings are used for qualification of seismic loads.

1.5.3 Overlapping l l 1.5.3.1 General Overlapping is used to separate seismically analyzed piping problems. Isolation of non-seismic piping I from seismic piping is addressed in Section 1.5.2.3 of this appendix. Seismic piping that cannot be separated by decoupling as described in Section 1.5.2 of this appendix may be separated using an overlap region. Where an overlap region is used, an adequate number of rigid restraints and bends in three directions to prevent the transmission of motion due to seismic excitation from one end to the other is included. The following criteria is used for applying overlapping: 3 NRC Staff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Approved Design Meterial- Design of SSC Page 3.9A.15

                                           -                                         4 CESSARMi h e,.

1.6.2 ASME CLASS 2 AND 3 PIPING The aAlowable stress limits for the specified loading combinations for ASME Class 2 and 3 piping are shown in Table 3.9-11 and Reference 4.12. l 1.6.3 ALLOWABLE NOZZLE LOADS l Equipment nozzle loads are minimized to be within equipment vendor specifications. The design nozzle load values are provided to equipment vendors as part of the procurement specification. 1.6.4 ALLOWABLE PENETRATION LOADS Piping systems are designed such that loads and displacements on ! containment penetration assemblies, as shown in Figure 3.8-2, meet manufacturer's allowables. 1.6.5 WELDED ATTACHMENTS ' Per ASME Section III, Subarticle NC/ND 3645, external and / internal attachments to piping are designed so as not to cause j flattening of the pipe, excessive localized bending stresses, or ' harmful thermal gradients in the pipe wall.puch attachments are designed to minimize stress concentrations in applications where l the number of stress cycles, due either to pressure or thermal ! effects, are relatively large for the expected life of the equipment. )

[ Local stresses due to all support loads acting on a welded attachment are evaluated and added dire

y to the nominal pipe /

stresses at the point of the attachment e sum of the stresses are compared against the allowable s resses given in Tables i 3.9-10and3.9-11.]%ethodsforevaluatinglocalstressesdueto welded attachments are provided in ASME Code Cases N-318 and N-392. Methods and criteria are supplemented by NRC-approved PVRC and EPRI testing and research. 1.6.6 FUNCTIONAL CAPABILITY REQUIREMENTS See Section 3.9.3.1.4.3. l 1.6.7 VALVE REQUIREMENTS l Piping systems are designed such that valve accelerations meet the allowable manufacturer's requirements for l seismic l acceleration. In lieu of specific values, reasonable generic seismic valve acceleration limits for SSE conditions and water hammer type loads are established. The design values are included in the procurement specification. The loads on supports Amendment Q 3.9A-21 June 30, 1993 l

 -.                                                                               1 c

CESSAR HSincam. I developed is a function of the coefficient of friction of the i sliding surfaces and the support stiffness in the direction of ! movement. Since friction iq, due to gradual movement of the pipe, ) such as thermal expansion,Mrictional forces are considered in # i

    , the support design under combined deadweight and thermal loading
    ' only.

Friction forces are applied in both directions of thermal expansion.y To account for these forces, the friction force. is calculated by ) using the smaller of CN or KX, where C is the coef ficient of friction and N is the component of force normal to the movement and K is the stiffness of the support in the direction of X. [Typicalcoefficientsoffrictionare: i 0.3 for steel to steel 0.1forlowfrictionslide/bearingplates]y l l Typically, frictional forces are neglected in the analysis of the l piping system because supports are designed to minimize the effects of friction on the piping analysis. 1.7.2.10 SucDort Gaps l Small gaps are provided for frame type supports built around the l pipe. These gaps allow for radial thermal expansion of the pipe as well as allowing for pipe rotation.[,. Total gaps of 1/8 inch or - less in the restrained direction are ne ligible and are considered to be zero in the piping analysis. 8' ) 1.7.2.11 Support Mass Typically, the mass of the support is not included in the piping analysis. Therefore, the weight of components supported by the pipe is limited to the extent possible. For example, spring supports include the weight of the components below the spring in the spring load setting, thus negating that part of the spring support weight that is supported by the pipe. However, due to the seismic response of the attached mass, supports which add substantial mass to the pipe are evaluated for the effects on the piping analysis. 1.7.2.12 Welded PiDe Attachments Welded attachments to the pipe wall are avoided where possible. However, certain design requirements such as anchors or axial restraints require the use of welded lugs or trunnions. All l welded attachments require the evaluation of the local stresses induced into the pipe. Materials used as welded attachments are compatible with the piping material. i Amendment P i 3.9A-26 June 15, 1993

ip- : C

5% System 80+ Design ControlDocument The LBB piping evaluation plot requires performing two complete LBB evaluations. The evaluations are l for two NOP loads which span the typical loadings for the line under consideration. A completed typical i diagram is shown in Figure 3.9A-28. The procedure used for generating that figure is as follows:

1. Choose NOP = Pressure + NOP, i

2. Determine ai I

3. Increase the :nalysis moment until the critical moment is found for ai and 2a,

4. Separate the critical analysis moment, M,, into the correct addition of SSE and NOP, proportion .

for the ai and 2a, evaluations. M, = g (NOP, + SSE i) (ai Analysis) M' and

                                      - NOP:

SSE = -# M, = (NOP, + SSE i ) , (2ai Analysis) SSEi = M, - NOP, /

5. Plot SSE values at NOPi for the ai 2ai analyses, respectively. This corresponds to the points labeled "1" in Figure 3.9A-K 28

6. Repeat steps (1) to (5) for NOP2 . The results are shown in Figure 3.9A-28, labeled "2". l l

Two stability evaluations are performed for each pipeline under consideration in order to complete the ! piping evaluation diagram. ' When stratified flow (SF) is a critical thermal transient that must be considered in the stability analyses, the PED is constructed using the following relationships for the ai and 2a, evaluations in (4) above: 1

4. M, = g SF i (a, Analysis) l l

l M l SFi = - I and d M, = SF i (2a, Analysis) SFi = M, As in the case for SSE, Step (4) is repeated to determine SF2 for the ai and 2ai analyses, and SF is plotted vs. NOP. Approved Design Meteniel Design of SSC Pepe 3.9A-32

l l System 80+ Design ControlD"cument 1 i e inconel 625

1. Swelling Available information indicates that inconel 625 is highly resistant to radiation swelling.

22 Exposure of Inconel 625 to a fluence of 3 x 10 m1 (E>0.1 MeV) at a temperatute of 400*C (725*F) showed no visible cavities in metallographic examinations (Reference 52) so that swelling, if any, would be very minor. Direct measurements made after exposure 22 ofInconel 625 to a fluence 5 x 10 mt (E > 0.1 MeV) at LMFBR conditions showed no 22 evidence of swelling (Reference 53). Further exposure to 6 x 10 nyt (E > 0.1 MeV) at 500*C (932*F) showed essentially no swelling as measured by "nmersion density, but did show small cavhier,. Thus, Inconel 625 is not expected to wil below fluences of l 3 x 1022 nyt (E > 1 MeV). l l 2. Ductility l The ductility of Inconel 625 decreases after irradiation. Extrapolation of lower f!uence data on inconel 625 and 500 indicates that the values of uniform and total elongation of laconel 625 after 1 x 1022 mt (E> 1 MeV) are 3 and 6%, respectively.

3. Strength The value of yield strength for Inconel 625 increases after irradiation in the manner
typical of metals. However, no credit is taken for increases in yield strength in the I design analyses above the value initidly specified.

l 4.2.1.5 Surveillance Program 4.2.1.5.1 Requirements for Surveillance and Testing of Irradiated Fuel Rods High burnup performance experience, as described in Section 4.2.3, has provided evidence that the fuel will perform satisfactorily under the design conditions. Two irradiation programs were developed for fuel performance surveillance in Arkansas Nuclear One-Unit 2 (ANO-2). The fuel rods in these 16x16 fuel assemblies are similar to those in the System 80 design. The first fuel performance program in ANO-2 has been completed. This program followed six standard assemblies through three irradiation cycles. Each assembly contained pre-characterized fuel rods which were examined during refueling shutdowns. The results of the program demonstrated that the fuel assemblies performed reliably through averaged burnupsof 37.2 GWd/MTU. Zircaloy oxide thicknesses, fuel rod growth and bowing, and assembly dimensional stability were consistent with initial predictions (Reference 54). l The second program at ANO-2 irradiated two fuel assemblies containing both standard and advane.ed design fuel rods to extended burnups. Both assemblies were extensively pre-characterized. One asserably was irradiated for three reactor cycles and reached an assembly-averaged burnup of 33 GWd/MTU. A second assembly was exposed to 5 cycles and reached an assembly-averaged burnu, of 52 GWd/MTU (Reference 55). Both assemblies were examined after each reactor cycle. Visual examinations, oxide thickness measurements, and other dimensional measurements result /gthe[ conclusion that the performance of the fuel has been satisfactory. D~*'uHve 9 ~" re '6e&!sto wught* th* "h*"**'d;:,s3 GI [G;i b& Yin' M kNhNC htk 4eb Approved Design Material- Reactor Page 4.2 29

i System 80+ Design ControlDocument }; 1 Table 4.3-1 Nuclear Design Characteristics Item Value ] General Characteristics

Fuel management 3-batch, mixed i central zone Core Average Burnup (MWD /MTU),10 ppm soluble boron 16,000
Core Average U-235 Enrichment (wt%) 2.6 j Core Average H2 0/UO2 volume ratio, first cycle, hot (core cell) 2.06 j

1 Number of control element assemblies -j k<4) Full strength 68 Part strength 25 i Burnable Absorber Rods m Number 11,680 r, Material Er2O3 , Io*N' orth $ 4, at BOC 4t,Oh f w <} g M , k8 4 4 7e o p d SOC, Gld, W " P 5.3

;                       y                                                                      4.0 j  -          Dissolved boron content for criticality, ppm, (CEAs withdrawn, BOC)

A-j Cold, 68'F (, 1431

,                           Hot, zero power, clean,        F                                  1414 Hot, full power, clean, 587'F                                     1270 i                            Hot, full power, equilibrium Xe                                   1006 Dissolved boron content, ppm, for:

f Refueling 2150 5 % suberitical, cold, first cycle (all CEAs out) O M TU 1837 1 5% subcritical, hot, first cycle (all CEAs out) 0 MTU 1920 Boron worth, ppm /% 4 (BOC/EOC)

!                                                                                                         w Hot, 587'F                                                                     96/91 Cold, 68'F                                                                     78/66 Neutron Parameters i            Neutron lifetime (Cycle average), microseconds                                  28.4 Delayed neutron fraction (cycle average)                                       0.0061 Plutonium Buildup (first cycle) e Fissile Pu (fmal)                                                             4.68
.             kg U (original) c Total Pu (final)                                                               6.02
,               kg U (original)

Approved Design Meterial- Reactor Page 4.3-33

System 80+ Desian ControlDocument Table 4.3-2 Effective Multiplication Factors and Reactivity Dataill Condition M p Cold, 68'F (0 ppm), BOC1 1.233 0.189 Cold (68'F) at minimum refueling boron concentration (2150 ppm), 0.916 -0.092 BOC1 5$le Hot,6S7'F, zero power, clean (0 ppm), BOC1 1.173 0.148 i Hot, full power, no Xe or Sm, 587'F (0 ppm), BOC1 1.148 0.129 l Hot, full power, equilibrium Xe (0 ppm) 1.111 0.100 Hot, full power, equilibrium Xe and Sm (0 ppm) 1.107 0.096 Reactivity decrease, hot Zero to full power, BOC (911 ppm) 0.014 Fuel temperature 0.012 Moderator temperature 0.002 Reactivity decrease, hot Zero to full power, EOC (0 ppm) 0.020 Fuel temperature 0.011 Moderator temperature 0.009 (1) No control element assemblies or dissolved boron except as noted, initial core. Approveef Design Materie! Reactor Page 4.3-34

i System 80+ Design ControlDocument i Table 4.3-4 Reactivity Coefficients i Moderator Temperature Coefficient,4/*F Value

Beginning-of-cycle (0-50 mwd /MTU)

, Cold, 68'F, Cleand431 ppm -0.20 x 104 Hot aro power, $P,'F, no CEAs, Clean,1400 ppm -0.03 x 104 i Hot full power, 587'F, no CEAs, Clean,1284 ppm -0.37 x 104 j Hot full power, 5g*F, no CEAs, Equilibrium Xe,1006 ppm -0.65 x 10d Hot aro power, .Wr'F, regulating CEA banks 3,2 and 1 inserted, -0.69 x 10 4 l 50 mwd /MTU,1006 ppm, Hot full power equilibrium Xe i Endef-Cycle (10 ppm soluble boron,16,000 mwd /MTU) Cold, 68'F (apprgte) -0.04 x 104 j Hot uro power,E57'F, no CEAs, Hot full power equilibrium Xe -1.70 x 10 4 . Hot full power, egbrium Xe, no CEAs, 587'F -2.60 x 104 q Hot aro power, 3FF, rodded, regulating CEA banks 3, 2 and 1 -2.00 x 104 4 insened, Hot full power equilibrium Xe Moderator Density Coefficient, 4/gm/cm' Hot, operating, 587'F Beginning-of-cycle,1284 ppm soluble boron, O mwd /MTU +.031 Fuel temperature contribution to power coefficient, 4/(kW/ft),1006 ppm, 50 mwd /MTU j Hot aro power -2.08 x 10-3 Full power -1.77 x 10-5 4 j Moderator void coefficient 4/% void Hot, operating, 587'F

Beginning-of-cycle,1284 ppm soluble boron, O mwd /MTU -0.22 x 10 8 Moderator pressure coefficient, 4/ psi Hot, operating, 587'F 2 Beginning-of-cycle,1284 ppm soluble boron, O mwd /MTU +3.% x 104 1

l Overall power coefficient, 4/(kW/ft)

Hot, operating, 587'F j Beginning-of-cycle,1006 ppm soluble boron,50 mwd /MTU -1.84 x 10-3 j End-of cycle,10 ppm soluble boron,16,000 mwd /MTU -4.31 x 10-3 i

1

3 Approved Design Material- Reactor page 4.3 36

__ . . - _ ~_ _ _ - . . _ . . _ . -_ . _ . _ _ - _ _ _ _ _. . _ __ _ _ _ __. _ E l System 80+ _ Deskrn contret Document Table 5.1.1-2 Per tar Coolant System Design Parameters Parameter Value Design Thermal Power, MWt 3931 , (including Net Heat Addition from Pumps) l Thermal Power Developed by the RCS, Btu /hr 1.342 x 10 ' Design Pressure, psia 2500 Design Temperature (except pressurimr), 'F 650 Pressuriar Design Temperature, 'F 700 Coolant Flow Rate, Ib/hr 165.8 x 10' Cold 12g Temperature, Operating, 'F $55.8 Average Temperature, Operating, 'F 585.4 Hot leg Temperature, Operating, 'F 615 Normal Operating Pressure, psia 2250 l System Water Volume, ft'(Without Pressuriar at full power) 13,625 I

Pressuriar Water Volume, ft' (full power) 1200 . Pressurimr Steam Volume, ft' (full power) 1234 Reactor Coolant System Vol Table 5.1.1-3 (at 70* Component Volume (ft') Reactor Vessel . 5,758 Steam Generators 2,973 (each) l l Reactor Coolant Pumps 134 (each) Pressuriar 2400 l Piping. ) Hot leg 135.6 (each) i Cold Leg 213.7 (each) Surge Line (nominal) 58.8 l l l OIen b Q ! OUdEi 7- e W w+ suo s as cnia co,i+ks es m 1>b><ruads Wk c W e'Stia. Won so u)ka a t% re u d <. ( cxurv a s. th i a bCM fu bit u e- and<rs % d w% % frassuetYa r @ W1

are- du.Clu np 4 proved Deskn Materiel RC5 emt Connected Syenome page 6. r.s

l System 80+ Desian control Document 5.2 Integrity of Reactor Coolant Pressure Boundary i i This section discusses the measures employed to provide and maintain the integrity of the Reactor Coolant l Pressure Boundary (RCPB) throughout the facility's design lifetime. The RCPB is defm' ed in accordance I with ANSI /ANS 51.1-1983. Included are all pressure contauung components such as pressure vessels, l piping, pumps, and valves which are:

1. Pan of the Reactor Coolant System, or j

2. Connected to the Reactor Coolant System, up to and including the following:

The outermost contamment isolation valve in piping which penetrates the containment;
The second of two valves normally closed during reactor operation in piping which does not penetrate the contamment.

5.2.1 ConapHa- with Codes and Code Cases 5.2.1.1 CompHa- with 10 CFR 50.55a ! The codes and component classifications are listed in Table 5.21 and are in accordance with the provisions of 10 CFR 50.55a. The ASME Code Edition and Addenda used in the design and construction l of th,ejeactor coolant pressure boundary components are specified in Table 1.84 and discussed in Section Obb 1.8L5f ASME Code Editions and Addenda other than those specified in Table 1.84 and Section 1.8 are used they will be idemified to the Commission by the COL applicant and shall have been endorsed by b uc O 10 CFR 50.55a.M ' 7_

                                                                                                                      .D< e-5.2.1.2            Applicable Code Case                                         NF CCL
                                                                                                                        ?~ '

Reactor Coolant Pressure Boundary components are fabricated in accordance with the ASME Code, Section III. The applicable ASME Code cases listed in Table 1.8-7 are utilized in the plant design and manufacturing. If code cases other than those specified in Table 1.8-7 are used they will be identified to the Comnussion by the COL ppplicant and shall have been endorsed by Regulatory Guides 1.84,1.85 or 1.147 as applicable 3 Code cases not endorsed by the above Regulatory Guides may be used with specific authorization from the Commission under 10 CFR 50.55a. 5.2.2 Overpressure Protection 5.2.2.1 Design Bases Appendix SA presents the design bases for sizing the overpressurization protection system. The loss of load transient which is used to size the primary safety valves is not intended to be used as a design transient for any other NSSS equipment. . 1 CoI tr1fu er un b on lle ese c) s u Dc, p m Wo d -eJzo a S s a 5.c, Amwoved Design Mataniel . RCS and Carmected Systerns Page.64* 57-(

l l System 80+ Design ControlDocument j l It is recogmzed that the Combined Operating License applicant may decide to modify the conservative

approach in determining LTOP enable and disable temperatures in order to take advantage of the wider operating margins allowed in the adoption of the Branch Technical Position RSB 5-2 defmition. This modified approach would necessitate the imposition of admmistrative controls to require that the technical specification P-T Limits not be exceeded above the BTP dermed LTOP enable and disable temperatures. dThe Combined Operating License applicant must verify that the site-specific material properties used in hakc the reactor vessel beltline and the as-deposited welds are consistent with the material specifications set forth in Section 5.2.3.1. If not, new plant-specific pressure-temperature limit curves must be derived and b (Mj used in a reevaluation of the LTOP enable and disable temperatures.gl 7 5.2.2.10.2.3 Equipment Parameters ' y j The SCS relief valves (SI-179, SI-189), used to provide LTOP, are spring-loaded liquid relief valves with Col sufficient capacity to mitigate the most limiting overpressurization event. Pertinent valve parameters are as follows: Parameter Value Nommal Setpoint 545 psia

Accumulation 10 %

Capacity 5000 gpm (@l0% acc) Since each SCS relief valve is a self actuating spring-loaded liquid relief valve, control circuitry is not required. The valve will open when RCS pressure exceeds its setpoint. The SCS relief valves are sized, based on an inadvertent safety injection actuation signal (SIAS) with full pressurizer heaters operating from a water-solid condition. The analysis assumes simultaneous operation of four SIS pumps and one charging pump with letdown isolated. The resulting flow capacity requirement for water is 5000 gpm. The analysis in Section 5.2.2.10.2.1 assumed that either SCS relief valve relieved water at this rate. The design relief capacity of each of two SCS relief valves (shown in P&ID Figure 6.3.2-1C) as supplied by the valve manufacturer meets the minimum required relief capacity of 5000 gpm which contains sufficient margin in relieving capacity for even the worst transient. The SCS relief valves SI-179 and SI-189 are Safety Class 2, designed to Section III of the ASME Code. Testing of valve relieving capacity and set pressure are discussed in Section 5.4.7.4. 5.2.2.10.2.4 Administrative Controls ! Administrative controls necessary to implement the LTOP provisions are limited to those controls

necessary to open the SCS isolation valves.

During cooldown, when the temperature of the RCS is above that corresponding to the intersection of the controlling P-T Limit and the pressurizer safety valve setpoint, overpressure protection is provided by the pressurizer safety valves, and no administrative procedural controls are necessary. Before entering l Pressure rneasured at the valve inlet. I f. 'OL. M 4 n u m u N9 u- su %h %n d., & th e, a 3. % 1 Approved Design Materini- RCS and Carmected Systems Po -1 S, z -/4 i

System 80+ Desogn ControlDocument Materials used in the reactor vessel beltline and the as-deposited welds contain no greater than the following weight percentages of residual elements: Copper (in welds) 0.03 Copper (in forgings) 0.03 Phosphorous 0.012 Nickel (in forgings) 1.00 Sulfur 0.010 Nickel (in welds) 0.10 Vanadium 0.030 -

   -                                                                                                                           l
  /  .The COL applicant will verify that the plant-specific material properties are within the limits assumed int
  -y   the CESSAR-DC and the information will be provided in the site-specific SAR.j                                      ,d 5.2.3.2          Comp.ulcay with Reactor Coolant                                                                             .jocd .

5.2.3.2.1 Reactor Coolant Chani=*ry pp oc Controlled water chemistry is maintamed within the RCS. Control of the reactor coolant chemistry is the function of the CVCS which is described in Section 9.3.4. Water chemistry limits applicable to the k - RCS are given in Section 9.3.4. W 5.2.3.2.2 Materials of Construction Compatibility with Reactor Coolant j, l The materials of construction used in the RCPB which are in contact with reactor coolant are designated l by an "a" in Table 5.2-2. These materials have been selected to mmunize corrosion and have previously l demonstrated satisfactory performance in other existing operating reactor plants. Metallic materials in contact with reactor coolant shall be restricted in cobalt content to as low a level as practical for all stainless steel or nickel base alloy components with a large wetted surface area. Cobalt based alloys shall be avoided exc pt in cases where no proven alternative exists. 5.2.3.2.3 Compatibility with External Insulation and Environmental Atmosphere All metallic insulation used in the plant is of the stainless steel reflective type, which mmmuzes insulation contammation in the event of chemical solution spillage. All non-metallic insulation used in the plant is designed to meet the requirements of Regulatory Guide 1.36, " Nonmetallic Thermal Insulation for Austenitic Stainless Steel." Complying with this Regulatory Guide assures that the nonmetallic insulation is designed in a manner which mmmuzes the potential for stress corrosion of stainless steel due to leaching of chloride or fluoride ions onto the stainless steel surfaces. 5.2.3.3 Fabrication and PrQ of Ferritic Materials 5.2.3.3.1 Fracture Tough-s 5.2.3.3.1.1 NSSS Components Fracture toughness requirements for Reactor Coolant Pressure Boundary components are established in accordance with the ASME Boiler and Pressure Vessel Code, Section III. Fracture toughness testing of base, weld and heat affected zone materials will be conducted in accordance with the ASME Code. Data from these tests will be available after the required testing has been performed and may be examined upon request at the appropriate manufacturing facility. 1 Col is Ern -% 8 .C% G ee- Ocp S E c- T1= u 5 . 2-romved oes+, nearerw - scs arod comeecsed syssoms rh T. 2 -tB

1 I l CESSAR iRWAcar s 1 ! I d' 9 discussion concerning prediction of irradiation effects and the j material surveillance program, s g l j Testing and measuring equipment for fracture toughness tests for 4 the reactor vessel, steam generators,- pressurizar, piping and l i reactor coolant pumps are calibrated in accordance with i Subarticle NB2360 of the ASME Code, Section III. l ? .

5.2.3.3.2 Control of Welding

? l 5.2.3.2.2.1 Avoidamos of cold cracking i l C-E complies with the recommendations of Regulatory Guide 1.50, j control of Preheat Temperature for Welding of Low Alloy Steel,

May 1973, as discussed below.

Paragraph C.1.b implies that the qualification materials are an 4, ; i infinite heat sink that would instantaneously dissipate the heat l input from the welding process. The qualification procedure consists of starting the welding at the minimum preheat {f) j temperature. welding is continued until the maximum interpass 3 temperature is reached. At this time, the test material is j permitted to cool to the minimum preheat temperature and the 3 welding is restarted. Preheat temperatures utilized for low alloy steel are in accordance with Section III of the ASME Code.

The maximum interpass temperature utilized is 500*F. ,

k The paragraph C.2 recommendation is considered an unnecessary j bg extension of procedures which apply to low-alloy steel ds, i i C # 8-meeting ASME code Sections III and IX requirements. recommendations of Regulatory Guide 1.50 are met by complying The d l Jtem with paragraph C.4. The soundness of all welds is verified by p 4.l,ldf ASME Code acceptable examination procedures.] 4 (pN } With regard to Regulatory Guide 1.43, major components are l fabricated with corrosion resistant cladding on internal surfaces exposed to reactor coolant. The major portion of the material l protected by cladding from exposure to reactor coolant is SA-533B I class 1 or SA-508, Class 2 or 3. Cladding of SA-508, Class 2 l forging material is performed using low-heat-input welding i j processes controlled to minimize heating of the base metal. 5 Low-heat-input welding processes are not known to induce underclad cracking. j 5.2.3.3.2.2 Regulatory Guide 1.34 l Regulatory Guide 1.34 recommends controls to be applied during j welding using the electroslag process. The electroslag process is not used in the fabrication of any RCPB components. i I Therefore, the recommendations of this guide are not applicable. Amendment D i 5.2-18 4 September 30, 1988 i !

4 i e

System 80+ Deslan ContmlDocument i
As a result of the above tests, a relationship was established between the carbon content of 304 l stainless steel and weld heat input. This relationship is used to avoid weld heat-affected-zone j sensitization as described below.

e Unstabilized Austenitic Stainless Steel , The unstabilized grades of austenitic stainless steels with carbon content of more than 0.03 % used for composras of the RCPB are 304 and 316. These materials are furmshed in the solution i annealed condition. Exposure of completed or partially-fabricated components to temperatures j ranging from 800'F to 1500*F is prohibited. i ._ Duplex, austenitic stainless steels comaining more than SFN delta ferrite (weld metal, cast metal, ) weld deposit overlay), are not considered unstabilized since these alloys do not sensitize, that is j 4 form a cominuous network of chromium-iron carbides. Specifically, alloys in this category are: }

CF8M, CF8 Cast stainless steel (deln ferrite l 5FN to 30FN, 8FN to 20FN for normal j operatir.g temperature above 500*F) 4

                                -  308,309                        Singly and combined stainless steel I                              (                                      ,                                & d ASN              N 312,316                       Weld filler metals (delta ferrite              5 9 C' " /

t controlled to 5FN-15FN deposited)~ { In duplex, austenitic/ferritic alloys, chromium-iron carbides are precipitated preferentially at the ferrite /austenitic interfaces during exposure to temperatures ranging from 800-1500*F. This j precipitate morphology precludes intergranular pewsions associated with sensitized Type 300 j series stainless steels exposed to oxygenated or fluoride environments. 3 i

Avoidance of Sensitization 3

j Exposure of unstabilized austenitic Type 300 series stainless steels to temperatures ranging from { 800 to 1500*F will result in ettide precipitation. The degree of carbide precipitation, or sensitization, depends on the temperature, the time at that temperature, and the carbon content. ! Severe sensitization is defined as a continuous grain boundary chromium-iron carbide network. j This condition induces susceptibility to intergranular corrosion in oxygenated aqueous 4 environments, as well as those commining fluorides. Such a metallurgical structure will rapidly j fail the ASTM A708 Strauss Test or the ASTM A 262 Practice E Test. Discontinuous j precipitates (i.e., an intermittent grain boundary carbide network) are not susceptible to intergranular corrosion in a PWR environment. 1 j Weld heat affected zone sensitized austenitic stainless steels (which will fail the Strauss Test,

.                          ASTM A 708) are avoided by careful control of:                                                           $4 2

e (O W W \ Weld heat input to less than 60 kJ/in 7 0 v'- i bC-

;                          e        Interpass temperature to 350'F maximum j

i e Carbon content t2 . O e -f */ aq,,y , [ Clevi d cde AJ 4

)         Approved Design Atatenfel RCS and Conneceed Systems                                                           Page 5-21
)

l l System 80+ oesen canaat Docanent The ISI Program consists of three subprograms as follows: i

The Component Inspection Program, which includes piping system welds, hangers, supports, internal inspection of pump and valve bodies and bolting. This includes ultrasonic testing in accordance with Appendix VIII of ASME Section XI, Edition and Addenda required by 10 CFR 50.55a.
The Pump and Valve Inservice Test (IST) Program, which requires operability testing of selected pumps and valves, and

The Hydrostatic Test Program.

4 L/ The development of the PSI and ISI program plans is the responsibility of the COL applicant and will , be based on the ASME Code Section XI, Edition and Addenda specified in accordance with 10 CFR 50.55a. For design certification, the NSSS designer is responsible for desigmng the reactor coolant pressure boundary for ac essibility to perform the preservice and inservice mspections. 1 Responsibility for designmg other components for preservice and p of the COL applicant. ' The COL applicant is also responsible for specifying the Edition of the ASME k Code Section XI to be used, based on the procurement date of the component per 10 CFR 50.55a. ASME Boiler and Pressure Vessel Code, Section XI, Subsection IWH has not been -formally approved gg and issued for use by the nuclear industry. After approval and prior to implementation of the System Ng 80+ pre-service inspection and in-service iaspection programs, the COL applicant shall t vipw the B&PV g Code Subsection IWH to ensure appropriate complice of these inspection programs., The provisions of 10 CFR 50.55a(a)(3)(i) and (ii) shall apply in cues where the System 80+ design does not practically accomm*** full adherence to Subsection IWH. 5.2.4.1 System Boundary Subject to Inspection The reactor pressure vessel, pressurizer, prunary side of the steam generator and associated piping, pumps, valves, bolting and component supports are subjected to inspection. 5.2.4.2 Air -- of Systems and Components to Provide Accessibility " Accessibility to equipment for mamtenance, testing, and inspection is a basic element of the System 80+ design process as indicated in Table 1.2-1 (Item A.2). The layout and arrangement of the plant provide adequate working space and access for inspection of specific areas of Code Class 1 components of the ! RCPB in accordance with IWA-1500. The Code Clas 1 components of RCPB subject to inspection are l those components defined by ASME Code Section III, Division 1. All Class I components will be designed for and be provided with access to enable the performance of ASME Code Section XI inspections in the installed conditions. Systems and components are designed such that design, materials, and geometry do not restrict inspections required by Section XI of the ASME Code. Listed below are the provisions for access for exammation of the RCPB:

           > . 's-             Reactor Vessel and Closure Head n                                                               .
        #M I               Ss           From Inside the Vessel:

All internals of the reactor vessel (which is an open structure offering insignificant L inpediment to access) are removable making the entire inner surface of the vessel i Coi w vm_4-rea -% ; W xD : c c n w 3 .2-

Approvenf C ,. Acetonio!- RCS amt Connecenef Systems Page 6-24

i i i t System 80+ Deslan ControlDocument ) l Table 5.2-2 Reactor Coolant System Materials i t j Cornponent Material Specification ,

                                                                                                                                                              '            I Reactor Vesseli'                                                                                         o .-                      f Forgings SA-508 Class 2OB Qadding*                                                           Weld deposited sustenitic stainless steel

{ - with 5FN-18FN delta ferrite or NiCrFe ' alloy (equivalent to SB-166) f 5 4 DVI Nozzle Safe Ends SA-182 F316 i Reactor vessel head" CEDM Nozzles NiCrFe Alloy 690 (SB-166) ; r-Ancrenitic Stainless Steel and CrNiFelalloy M ' "' Vessel internals" j i 4 Flow skirt" NiCrFe Alloy 690 (SB-168)

 ,                                                                                                                                                                          I

~ Fuel cinMmr* Zircaloy-4 l 1 j instrument nozzles

NiCrFe Alloy 690 (SB 166) 0 Control c!cinent drive mecharimm housingsT

                                                                                                                                                   .f lower"                                                             Type 403 stainless steel according to Code
!                                                                                          CCase N-4-11 with end fittings to be SB-166 )

i Alloy 690* and SA-182 Type 348 stainless steel 4 ) ]! Upper" SA-479 and SA-213 Type 316 stainless steel with cod fitting of SA 479 Type 316 and vent valve seal ; of Type 316 and vent valve seal of Type 440 stainless steel seat Oosure head bolts SA-540 B Pressurtzer : [ rum. .jp /, . k -/ 4 Shell S 533 G > Cass 1 or SA-508 Class 3 4 } Cladding" el austenitic stainless steel with D^ i 5 FN-18FN delta ferrite or NiCrFe alloy j j (equivalent to SB 166) l Forged nozzles [ SA-541, Class 3 or SA-508 Cass 3 q

Instrument nozzles and heater sleeves * / NiCrFe Alloy 690 (SB-166) _ f f -

y j Nozzle safe ends * / SA-182, F316 or F347 / 4 I / Safety valve nozzle flange" / SA-182 /

;                         Studs and nuts                                                      SB-637                                                               .

Steam fenerator 2%

P, n M d ^' h 2 L Qw h a \ [ Sc;< ts, c by 6 b ry ps bW " Y j Rejes c a fe h s w f (?n- 1960 % i$ )S w

                                                                  *is.                     ,       m              m            % rwe s Page 5-36 j         Aweved Design Marwiel RCS amt Connected Systems g g n p {

I

,/

System 80+ w ~ Desion control occummut 1

                            .-                                                  x                                                                                       g Table 5.2-3 ASME Code Cases i                     N-4-l l                               Requirements for Special Type 403 Modified Forgings and Bars l

l N-411-1 Alternative Damping Values for Response Spectra Analysis for Class 1,2 and 3 Piping Section II, Division ! N-71 15 Additional Materials for Subsection NF, Classes 1, 2, 3 and MC Component Supports Fabricated by Welding, Section III, Division 1.

                   /NU                is/ - .'- o- 4       Matenal for Core Suppon Structures, Section III, Division 1.

N-474-1 Design Stress Intensities and Yield Strength Values for UNS N06690 with a mmimum specific Yield Strength of 35 KSI, Class 1 Components, Section II, Division 1. /

                                                                                                                                             /
                                                                                                                                           -                /9 N                        Code Cases intended for use are in accordance with Regulatory Guide 1.85,(Revision 287
                \ ote:                     *Matenals Code Case Acceptability ASME Section III Division 1,"and Regulatory Guide 1.84(f s
                                          " Design and Fabrication Code Case Acceptability ASME Sectionlli, Division if
                                                                                                                              /          I
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                                                                                @ . $Og                                                              "           q bo d g h .m h                      'M)"d-s Cc 64 4 cod e Approved Design Meteniel- RCS ared Cormected Systems                                                                                 Page 5-40

          .,n.   .ss . . -         e .w     ,                      ,.u- .         a   a   .w     ~        - - - - - -                . . . . . _ . ,

4 P s System 80+ Deslan ControlDocument I j

Regulatory Guide 1.71, Welder Qualification for Areas of Limited Accessibility, is addressed in l Section 5.2.3.3.

e Regulatory Guide 1.99, Radiation Embrittlement of Reactor Vessel material, is addressed in Section 5.3.1.6.7.

I 5.3.1.5 Fracture Toughness i In accordance with 10 CFR 50 Appendix G, Paragraph IV A, the reactor vessel beldme materials have a minimum upper-shelf energy, as determmed from Charpy V-notch tests on unirradiated specimens in j accordance with Paragraphs NB-2322.2(a) of the ASME Code, of 75 ft-lbs. Charpy impact tests will be

performed on transversely (weak direction) oriented specimens from the beltline forgings to establish: RT,sor as required by 10 CFR 50, Appendix G.

j 5.3.1.6 Reactor Vessel Material Surveillance n y,s.gii i De irradiation surveillance program for System 80+ will be conducted to assess the neutron-induced J 4 changes in the RT,,or (reference %.mre) and the mechanical properties of the reactor vessel j materials. Changes in the impact and mechanical properties of the material will be evaluated by the j comparison of pre- and post-irradiation test results. De capsules containing the surveillance test ]1 specimens used for monitoring the neutron-induced property changes of the reactor vessel matenals will j be irradiated under conditions which represent, as closely as practical, the irradiation conditions of the l reactor vessel. ASTM E-185-82, Practice for Conducting Surveillance Tests for Light Water Cooled Nuclear Power Reactor Vessels, and 10 CFR 50 Appendix H, Reactor Vessel Matenal Surveillance Program j (9

;                           Requirements, present criteria for monitoring changes in the toughness properties of reactor vessel beltline j                           matenals through surveillance programs. The System 80+ reactor vessel surveillance program adheres

4 , to all of the requirements of ASTM E-185-82 and satisfies 10 CFR 50, Appendix H requirements.

j c j , 5.3.1.6.1 Test Material Selection l" ji-Materials selected for the surveillance program are thosejudged most likely to be controlling with regard to radiation embrittlement according to the recommendations of Regulatory Guide 1.99, Revision 2. lh j, Surveillance test materials are prepared from the actual materials 'ised in fabricating the beltline region of the reactor pressure vessel. He test materials are processed so that they are representative of the 1 materials in the completed reactor vessel. Specimens are prepared from three metallurgically different I 3

]d n
!'             N(materials, including base metal, weld metal and heat-affected                            TN9: zone (HAZ) material.
    * . QO      .t          In addition, material is included from a standard heat of ASME SA-533 Gi+ade B Class 1 manganese-I molybdenum-nickel steel made available by the USNRC sponsored Heavy Section Steel Technology 2                      (HSST) Program. His standard reference material (SRM) is used as a monitor for Charpy impact tests,
 't                    ( permitting, comparisons among the irradiation _ data from operating power reactors and experimental _ n
! $. ,                 \ rea_cWn. Compilation of data generated from post-irradiation tests of these correWtion monitors will be ~

d:arried out by the HSST_ program. _ ___ .__ l$! Base metal test material is from a section of the shell course forging selected from the beltline of the reactor vessel. Selection shall be based on an evaluation of initial toughness (characterized by an index

;y                          temperature such as RTm) and the estimated effect of chemical composition and neutron fluence on the i                           Appnuvenf Design hieterial . RCS anaf Cannectent Systenee                                          Pere 6.3 3 I.                                                                                         -..                      _                              _

l l l System 80+ oesten controlDocumeur i M, = Bending correction factor. E, ,

                                            = M, 6 /a/Q where, Mg and M. = correction factors defined by Figure A3-1 and A3-2 in Appendix 3 to WRC Bulletin 175.

a = crack depth, in.

Q = the flaw shape factor defined in Appendix 3 to WRC Bulletin f 175. K = Stress mtensity factor for bendmg stress due to pressure, ;

                                              ='o[,M, v
                                   ,[a      , = Secondary bending stress due to pressure, psi.

V d K = (1-p)(37) l l where, & = Stress intensity factor for thermal stress, cr = Coefficient of thermal expansion, in/in, I E = Modulus of Elasticity, psi, l p = Poisson's ratio, , AT = Temperature differential causing the stress, *F. I

right Q(2xy The + xQlefts side of theasabove calculated a function equation (Ka) is of pressure. Thecalculated max for vario

              #           allowable pressure corresponding to a given temperature can then be calculated from this j                          equation. The resulting (maximum) pressure-(nummum) temperature coordinate:: define l                          the flange PT Limit for a specific heatup or cooldown, at a particular point in p'. ant life.

l In no case is the flange minimum temperature limit allowed to be less than the flange RTm + 120'F during normal operation (when the pressure exceeds 20% of the preoperational system hydrocratic p=~-e by RT + 90*F during hydrostatic @OM pressure tests and leak testsCor by RTm + 60*F when the core is critical in accordance with Appendix G to 10 CFR 50. .

                                                                                                           ]                      4-
                                                                                                                                 ~2.

6

                                                      ) O C. f= C- 50       5 h 4-a             ThaA-      (r.-r g , + s o

c fo e LG )

(ort Clu os q pp h t a b l# to g u) (2,g go f Pu (2_- Approved Design Meteriel RCS emf Cormacted Systems Pope 5.315 i. i

am g aKC NOZZLE SCHEDULE SERVICE CEDM UPPER COOLANT INLET 4 ASSEMBLY COOLANT OUTLET 2 CEDM AND INSTRUMENTATION NSTRUM NTATION ic3 NOZZLES %CEDM MOTOR NOZZLES C Mini %.j m HOUSING ASSEMBLY IN-CORE INSTRUMENTATION 61 % E I I I . o VENT-RCGVS 1 1

                                         '          }1;b                 M:                         CLOSURE SEAL LEAK MONITOR               1                f                                                    STUD (54)

V NT-RCMS DIRECT VESSEL INJ. 4 , MATING SURFACE SEAL LE AK MONITOR I: 34.,7 :f 12' NOM. , j yHEATED JUNCTION DVI NOZZLE % % THERMOCOUPLE 16'-4.32' (8.5* ID) PROBES .j [- tl ,'1

                                                       ,     F ^"r'"r".'           ,

( 3r ID ,, .9)mena .fa) 42* ID

                                           ,j     .
                                                       ,                                             OUTLET INLET                1/     l
                                                    ]                                                NOZZLE m
                      * *** [ Q     t Ol CORE EXIT THERMOCOUPLE +

48' RASE METAL THICKNESS IN CORE REGION 91/16'--* +' . CLADDING 1/8* MIN '

CORE

                                             ,               DIAMETER                                          39'-1.35' 15'.2 1/4'
                                       ~

{ *- SNUBBER CORE STOP FLOW BAFFLE

                                       $                                               l NCORE INSTRUMENT NOZZLES                                                    -
                                                             .                                     Amendment U - 12/31/93 REACTOR VESSEL

1 l 1 System 80+ Design ControlDocument 5.4.10 Pressurizer

                                                                                         @ ed # %                     R f             f-5.4.10.1          Design Bases                                                        a c.@     f74 4 d The pressurizer is designed to:                                        @s/ (o d 6^

Maintain RCS operating pressure such that the minimum pressure during operating transients is {f above the setpoint for the Safety injection Actuation Signal and low pressure reactor trip, and Y such that the maximum pressure is below the high pressure reactor trip setpoint.

                                                                                                                                              'e
   -

Withstand the consequences of the design thermal transients of Table 3.9-1 without failure or 4 ,

malfunction. , g' l v ' i o

Provide sufficient water volume in the pressurizer to prevent uncovering the heaters as a result a of a reactor trip. 'y
Provide sufficient water volume to prevent pressurizer' heaters from being uncovered by the 'y ,

outsurge following step load decreases of 10% starting within the range of 100% to 25 % of full- c t rated power or a 5% per minute ramp decrease from 100% to 15% of full-rated power. 4 j d Q. '

Provide sufficient steam volume to avoid lifting the primary safety valves as a result of a loss of * ] t/) )

condenser vacuum (LOCV) event (normal control systems are operational) g , Y , o Provide sutticient steam volume to allow acceptance of the insurge resulting from any loss ofload [.I transient without liquid or two-phase flow reaching the prunary safety valve nozzles. .

Minimize the total reacter coolant mass change and associated charging and letdown flow rates -4 0 I in order to reduce the quantity of wastes generated by load follow operations. 5 Q

i ,d

Provide sufficient pressurizer heater capacity to heat up the pressurizer, filled with water at the ag zero power level, at a rate that ensures a pressurizer temperature (and thus pressure) which will _y maintain an adequate degree of subcooling of the water in the reactor coolant loop as it is heated =

by core decay heat and/or pump work from the reactor coolant pumps. $.

                                                                                                                                            -o l

Contain a total water volume that does not adversely affect the total mass and energy released to c.e 3 l the containment during the maximum hypothetical accident. .

l ut . ty a is- m"5 -

Ensure that,in addition to beiry/specified as Seism /ic Category I, the pressurizer vessel, including heaters, baffles, and supports /s I be desi nedluch that no damage to the ugggnLis ca s and -1 cps. The lower frequency is efm' ed as for tIi2f,n rq@by the ctorfrequency vehs ranges of 9-2The design basis for the higher frequency consis which diminishes internally within the vessel. 8 h @ f-o

                                                                                                                                   %                            ~2 The combination of maximum heat loss from the pressurizer and the pressurizer heater capacity                                                            i shall be such as to maintain the pressurizer at normal operating pressure during hot standby conditions. This capability shall be provided by redundant trains of heaters powered from off-site                                                      j power and Class IE emergency power.

I o The total spray flows shall be sufficient to keep the pressure below the reactor trip setpoint during l an insurge of water during the " Maneuvering and Load Follow" and " Loss of Load" transients. 4 proved Deelen AceterW RCS and Commcted Systems Pope 5.4-28 l 1

                                                            .                  . . _ . -                                    . _ ,      .                             .1

System 80+ Deslan controlDocument Table 5.4.10-1 Pressurizer Parameters Property Parameter Design pressure, psia 2500 Design temperature, 'F 700 Normal operating pressure, psia 2250 Normal operating temperature. *F 652.7 Internal free volume, ft) 2400 Normal (full power) operating water volume, ft 3 1200__ [ N l Normal (full power) steam volume, ft) f20j0 g Z 5'4 / Installed heater capacity, kW 2400 x 'gg l Heater type g immersion ' Spray flow, minimum design capacity, gpm 375 ' n Bypass Spray flow, continuous, gpm 1-6 l30 C Nozzles Surge, in. (nominal) 12, schedule 160 b Spray, in. (nominal) 4, schedule 160 Safety valves, in. (nominal) 6, schedule 160 Safety deprer;unzation, in. (nominal) 6, schedule 160 Instrument level, in. (nominal) 3/4, schedule 160 l Temperature, in. (nominal) 1, schedule 160 i Pressure, in. (nominal) 3/4, schedule 160 Heater. 0.D., in. 1 1/4 - ceMer i d M=- Buc M 2 C ke< e fo be cas a 4 4 ce i Me T46u s . i . i - z. i Apnproved Design Meteniel- RCS amt Connected 5yseems page 5.4 45

I

! System 80+ Design control Document Systems susceptible to an ISLOCA are to be designed so that all of the following conditions are satisfied without any operator action... I e the system retains its structural integrity throughout the event (structural integrity is preserved i if, by definition, the system maintains its pressure boundary despite distortion and/or loss of function); j j 4 e any leakage caused by the event is limited to the makeup system capabilities; and e offsite doses are limited to a small fraction of those specified in 10 CFR 100 as is assumed in the i design bases for the Chapter 15 analyses. 3.3 Compliance Methods Design responses to ISLOCA challenges discussed in this report constitute 1 i e increasing the design pressure rating of equipment or systems, Option A, and

i ! e incorporating design features which terminate and limit the scope of the ISLOCA event, Option j B. Option A design features rely on inherent physical attributes of a system or subsystem which will prevent ! I failure when it is pressurized to normal RCS operating pressure. Option A features do not require any .

!                      immediate action by equipment or operators to satisfy the ISLOCA acceptance criteria. This approach

is intended to provide the optimum protection against ISLOCA challenges and to allow the o tor the bcl

,                      necessary time to properly assess and restore the system to normal conditions.{ Examples of tion A                       g i'                   [ features satisfying the ISLOCA acceptance criteria include:] Y
                                                                                                                                          -e'-

E e locating the system or subsystem completely within containment; hw e designing the system or subsystem to normal RCS design pressure; f4W yg j e (designing the system or subsystem to a pressure of at least 40% of the RCS normal pressure.

                                                                                                                                              #Dl D

i Austenitic stainless steel piping will rae a minimum wall thickness corresponding to stanprd Q! ' l weight for sizes less than 16 inch NPS and schedule 40 for 16 inch NPS and larger sizes]and 1 j e physically separating the system or subsystem from the RCS daring conditions when the RCS

pressure exceeds its design pressure.

I J Option B design features are design responses to ISLOCA events consisting of specific equipment and

 ;                     instrumentation which perform actions to prevent or mitigate the consequences of an ISLOCA. Option j                      B design responses that have been considered will not require operators to prevent or mitigate the event, but will eventually require operators to perform remedial action, inspection of equipment following the
-                      event and returning the plant systems to normal configuration.

j Option B design features are intended to be applied to systems for which it is impractical to apply Option

j A design features. i e 4 ~ Approved Desogn Materiel- RCS and Carmected Systems Page SE-3 5

, System 80+ Design Control Document l i then may clear as reactor power decays due to the reactor trip. Therefore, when a N-16 radiation , monitor detects the high gamma radiation condition, the alarm will be latched. Acknowledging the alarm i

;       does not reset the latch, the alarm latch must be reset separately by the operator. This logic is presented               l by Figure SF-96.

l { 5.6.3 Primary-to-Secondary Leakage Monitoring Program 4 The instrumentation described above will be incorporated by the COL applicant into a prunary-to- _ secondary leakage monitoring program that addresses three specific scenarios: gy 2

j Low Level and/or Slowly Increasing Primary-to-Secondary Leakagfe g
Rapidly Increasing Prunary-to-Secondary Leakage (as described in IN-91-43 and IN-88-99Eand g }-
Steam Generator Tube Rupture (No Leak Before Break)'. '

V f) ud'AI ' 1 The program will alto address instrumentation setpoints and methodology for equipment (including N-16 1 . monitors) used to detect steam generator tube leakage and ruptures commensurate with the above ~

scenarios. Jy 5.7 Reactor Coolani Gas Vent Systeur b^ b* /J i j 5.7.1 Description - fusic/roA/

The System 80+ design includes a safety grade Reactor Coolant Gas Vent SymentRCGV/) used to vent i non-condensible gases and steam from the reactor vessel upper head and the pressurizer steam space.

The system is comprised to two independent piping trains, one from the reactor vessel and one from the l pressurizer each with parallel valve paths in each train allowing for single failure operation. The l parallel path ficm the pressurizer and the reactor vessel upper head contains two isolation valves in series. i The pipe trains are routed to the IRWST and the Reactor Drain Tank (RDT). The valves are remote } manually controlled. f. j 5.7.2 Operation fru,.e.ho H' ) The RCGV/'is designed to be operable durirgall $pjp basis events. In the event a void forms in either

!      the reactor vessel upper head, the RCGV/ fnay"8e used to depressurize the RCS and remove the void in I

the RCG up[per head. valves to vent In depressurize steam and the pressurizer, if pressurizer the reactor coolant system. spray is not availab 5.7.3 Instrumentation Pressure indication is provided between the isolation valves in each RCGV/ train which alarms and provides the operator with indication that a valve did not reseat or the valve leaks. Temperature indication is provided downstream of RCGVfisolation valves to monitor leakage from the reactor coolant pressure boundary. 1 1 2 Nl l*1k W bOH E lC 6 t-Q. .D C- D B YOC e ben SC C. 'h

N 5. ?_ .

l Approved Design Material- RCS and Connected Systerns Pege 5F.39

System 80+ Desian control Document Table SF-4 Sequence of Events for SGTR e CEPAC Cents Event /' Steady-state 100% power 0 0 10 10 Initiation of SGTR'(one tube) 80 80 BackuptIters on and full charging gsc(c.c 495 480 Meactor Trip on subcooling snargin 7 500 500 / Main feedwater Inanually isolated y 535 M 20 SIAS actuated charging to Inax flow, --

                                           /                         letdown isolated                                           ,

680 ./ 690 Trip two RCPs at 1300 psia 1170 / 1140 h feedwater is automatically actuated with flow diverted to both steam

                   ,                                                 generators
                 /                                          s t'.fnt. C TAC l
        !                        -ras       n.b te       M n b ~ tt 3            wic_

j - bo va( o [ ' T AB /r k v'ieqiI l Po 4.< 61 6odad 1

                                                                                                    \(CfI                     'd A ($ O
                                                                                 'Fiq see s           A a ct Se-chow             IN        Cl~

fef.ees~9 lv (c Pu /u n ( o M[)s < t s . J Approved Design Material RCS and Connected Systems p,y, 5p49

1 CESSARENGem, gl . 4 2. t v. y l to boil. As steam in the line expands, this feedwater inventory is pushed into the steam generator and is boiled off by primary to secondary heat transfer. The expansion of the feedwater inventory into the affected steam generator has been considered in the analysis. The expansion is assumed to be isentropic. The isentropic expansion of the feedwater downstream of the MFIV is determined in SGNIII. As the affected steam generator depressurizes, the feedwater expands. At first the feedwater is i sub' cooled and the fluid which expands into the affected steam ' generator is pure liquid. Once the steam generator pressure drops below the saturation pressure of the feedwater, flashing ! starts to occur and then the fluid which expands into the steam generator is two-phase. The equations for each phase of the isentropic expansion process are given below. Subcooled s y(Pgg, T) =s 0 Myyv (P gg, T) =V M M

    ,,      t             - At s .r mh = mh l             )

SG' Saturated Mffv (Pgg) +Mv g g (PSG) =V Msf f(PSG) +Mso q(Pgg) M

                                    =s 0 Mf+       g f       9             f       g t - At Q t)

Mffh (Pgg) + Mgq h (Pgg) - N"* Mf+ Mg symbols M Feedwater mass, lbm V Volume of feedwater downstream of MFIV, ft 3 P gg Pressure of anected SG, psia T Feedwater temperature, 'F h Specific enthalpy, Btu /lbm v Specific volume, cubic feet /lbm Amendment G 6.2-27 April 30, 1990

System 80+ Desinn ControlDocument t

3

          -e          The component cooling water side of the CSS heat exchanger is designed to ASME B&PVC j                      Section III, Class 3.

6.2.2.2.5 System Belishility Considerations l I System Reliability is discussed in Section 6.5.2.5 l j 6.2.2.2.6 System operation

A-
System Operation is discussed in Section 6.5.2.6. ,

i ! 6.2.2.3 Design Evaluation 4 p

The CSS uses the Lechler (SPRACO Company) 1713A nozzlg, which provides a drop size distribution j which has been established by testing and found suitable for thB fission product removal function. The: CSS prov .' s a nozzle pressure differential of 40 psid which fixes the drop size distribution. The mass

! mean drop size produced at this differential pressure is conservatively assumed to be 1000 microns for l evaluating system performance. ! 6.2.2.4 Preoperational Testing i j Preoperational tests are conducted to verify proper operation of the CSS. The operational tests include

calibration ofinstrumentation, verification of adequate pump performance, verification of the operability i

of all associated valves, and verification that the spray headers and spray nozzles are free of obstructions. In addition, a preoperational hot functional performance test is made on the lnstalled CSS heat exchangers j as part of the precore hot functional test program. ! The CSS also undergoes a series of preoperational hydrostatic tests conducted in accordance with Section ) III of the ASME Boiler and Pressure Vessel Code. 1 j Tests of individual components or the complete Contamment Spray System will be controlled to ensure

that plant safety is not jeopardized and that undesirable transients do not occur.

t

6.2.3 Annulus Ventilation Systeen i

j The Annulus Ventilation System'(AVS) serves the space between the primary contauunent and the i secondary contamment. The system does not perform any normal ventilation function. However, it does i provide additional assurance against the release of radioactivity to the environment; therefore, it is i designed as an engineered safety feature and should be capable of operatmg and performing its function j durmg startup, power operation, hot standby and hot shutdown. The AVS has no effect on reactor l criticality and is designed for public and station personnel radiation protection only. j 6.2.3.1 Design Bases The purpose of this system is to produce and maintain a negative pressure zone in the annulus. This

!          mitigates the consequences of airborne products of radiation that might otherwise become an j           environmental hazard during and following an accident.

} 4 i Approwed Dee@n neeenriel Engheered Senoty Feemene Pege 6.2 27 l

                -    s-..                                            -                   ,                                     . . - -

  - - . - = __ _ - -                         - . -          .._ -- -               .. .-- .                    --.        . .         ._.       ___   -

) System 80+ "" S Centrol Document

q - y*> @ . Renee manual control of the automatically actuated containment isolation valves is provided. 2 < Automatic valves are installed in lines that must be immediately isolated after an accident. Those lines i which must remain in service after an accident have at least one remote manual valve. 1 I ] The integrity of the isolation valves system and connecting lines, under the dynamic forces resulting from j inadvertent closure while at operating conditions (e.g., main steam lines) is ensured by the performance of static and dynamic analysis on the piping, valves and restraints. 1 I The supports and restraints are applied such that integrity is ensured and pipe stresses and support i reactions are within allowable limits. Valves, in nonsafety-related systems where function permits, are ! nonnally positioned closed to minimize any release following a design basis event are equipped with valve } operators to move the valve rapidly. 4 j Containment isolation valves, operators and associated instrumernation and controls are environmentally l qualified for the normal and post-accident envirt,nmen's in which they are located as discussed in Section 3.11. 4 i 6.2.4.3 Safety Evaluation The containment structure and the containment penetrations form an essentially leak-tight barrier. i Allowable leak rates from the containment under design pressure condition are discussed in Section 6.2.1. ! Testing provisions and performance are also discussed in Section 6.2.1. Whenever practicable, isolation ! valves outside containment which are normally open and required to close on a signal to isolate the containment are designed to fail closed. ] In order to ensure that no single, credible failure or malfunction will result in loss of isolation capability, the closed piping systems, both inside and outside the containment, and various types of isolation valves provide a double barrier. The isolation valve and actuators are located as close as practical to the containment and protected from j missile damage. This minimizes the potential hazards that could be experienced by the system. \ i The integrity of the isolation valve system and connecting lines under the dynamic forces resulting from i inadvertent closure under operating conditions is ensured, based upon required static and dynamic ! analysis. a The supports and restraints are applied such that pipe stresses and support reactions are within allowable l limits a defined in Section 3.9.3. a A single containment isolation valve will be provided in each in-containment refueling water storage tank l (IRWST) suction line to the engineered safety features (ESP) pumps. 'the valve need not be enclosed i in a leak-tight enclosure, since all requirements of Note 56.1 in Appendix A of ANSI /ANS 56.2-1976 , are met as outlined below: f 2.-5A j e The valves are attached to lines which are extensions of co y V nt and enclosed in a pu r- 'e'

i room adjacent to the containment which l'as provisions for vironmental control of any fluid { leakage. The IRWST suction lines are shown in -

                                                                                              'r.. ;.2-0. The containment isolation valves i                                for these lines are located in the reactor building subsphere along with the ESF pumps and heat exchangers as shown in Figure 1.2-               e reactor building subsphere has provisions for i                       Approved DesQn 60etodel

ingkreeraf Sekty Feehues Page 6.2 37

}. k

                                ,--     w         <                    < ,. ,                        _-               ,       -.

n

i 1 ]

1 / A

System 80+ / Deslan ConsrolDocument ~~. l i

I i j environme control of any fluid leaksge (i.e., leakage is collected in the sumps shown in i Figure 1.2 , and the reactor subsphere area is maintained at a negative pressure, and all potential 6 radioactive releases are filtered and monitored as discussed in Section 9.4.3). y l lL A d no containment atmosphere can impinge d e The lines from the IRWST are always submerged upon the valves (see Figures 1.24,1.2Jf, 1.2 ). (C' l 3. 3 I.1.-5A j e The systems which the lines from the IRWST connect to outside containment are closed systems d j meeting the appropriate requirements of closed systems in the preceding standard (N271-1976), including 3.6.4 and 3.6.7 (see Figures 6.3.2-1 A and IB). , i A 1, e The valves provide a barrier outside containment to prevent loss of IRWST water should a leak a

!                      develop. The valves are closed remotely from the control room and the operator can detect j                       which one of the four valves failed to close by indication of high sump water level in the sump loc.ued within the quadrant in which the associated ESF equipment and valve is located (see                                     /

j Figua 1.2-Xshowing quadrant separation, ESF un+ m location and sumps). i A single containment isolation valve is provided in each of the four SIS direct vessel injection lines and i in each of the two parallel SCS discharge lines and SIS warmup bypass lines. A single isolation valve j is acceptable because these lines are in engineered safety features (ESF) systems and meet the specific i criteria of SRP 6.2.4, Section II: } l e The System 80+ SIS /SCS provides redundant containment isolation barriers with a single CIV. !. The two barners are: 1

1. The MOV isolation valves in the lines outside containment (or the water seal provided

) by the IRWST, as described in Section 6.2.4.2). These valves can be remote-manually j controlled from the Main Comrol Room. Valve position indication is provided in the 4 Main Control Room. 1 l 2. The closed system outside containment. As shown in Figures 6.3.2-1A, B, and C, the SIS forms a closed loop with the containment. I e The closed SIS outside containment is protected from missiles (see Section 6.3.1.2.3), is designed i to Seismic Category I standards and classified Safety Class 2 (see Table 3.2-1), and has a design j temperature and pressure of at least 350'F and 900 psig (see Figures 6.3.2-1 A, B, and C), which exceed the containment design internal pressure and 6.w.LJe in Table 6.2.1-3. i e t

The integrity of the closed SIS outside containment is maintained during norral plant operations.

The system is periodically pressure tested in accordance whh the pre-service and in-service inspection programs required by 10 CFR 50.55a and the ASME Code, as described in Section 6.6. The integnty of the system is checked during in-service testing of safety related _ $ pumps and valves, as described in Section 3.9.6. [ e A single active failure can be aw,,- -f=-f with only one isolation valve in the line: if the j active failure is the MOV failing to close, the Si pump flow maintains a barrier on the valves, j If the active failure is the Si pump failing to start, the IRWST water level maintains a barrier on j the valves, as described below. ii ApproventDesgn Ateneriel : 1SenotyFeennen Pege 6.2 38 j l !i l 1

Syst:m 80+ __ Table 6.2.4-1 COntaintneh! NIdtiSE VAln ddd Id5dlBr Odls $5Bld.) (Note 16) (Note i) I acen Relative Nw Direction to Relative to Valve Item Valve Egure Valve Costain. r==saw Arrangesment No. Service No' No. Type meat (CDC) Nonnel _? 4 Main Steam Line #2 from Seam 50-151 10.1-2 Glok Outside Out i O Generator F2 SG-558 Safety Outside (57) C (Note 18) SG-559 Safety Outside C ,. S0-560 Safety Outside C SG-561 Safety Outside C SO494 Safety Outside C 50-179 Globe Outside C SG-10g Gase Outside LO SG48$ Globe Outside C SG-182 $,,. Gem Outside C EF-23p 10.4.9-1.2 Gate Outside LO 5 Main Feedwater to Downcomer 50-172 $0.h 2 Gate Outside in 10 0 Nonle Steam Generator #1 SG-130 Gate Outside O (57) SO442 Check Outside O SO452 Circk Inside 0 6 Main Feedwater to Downcomer SG-175 50.k-2 Gem Outside kn 10 0 Nozzle Steam Generator #2 SG-135 Gase Outside O (57) SG443 Check Outside O SO453 Check Inside O 7 Main Feedwater to Economizer SG-132 10.k.2 Gate Outside In 2, O Nozzles for Steam Generator #1 SG-174 Gase, Outside O (57) SG-598 - Check laside i sO - 8 Main Feedwater so Economizer 50-137 10.1 2 Gam Outside In 2 O Nozzles for Steam Generator F2 SG 177 . Gam. Outside O (57) SG-599 Check Inside -O.

9 Motor-Driven EFW Pump #I EF 102 10$5 k.1 Gase Outside in 4 C Discharge EF-202 Check Inside (57) C 10 Motor-Driven EFW Pump F2 EF-10) Idi.9-1.1 Gau Outside In 4 C Discha.ge EF-20) Check Inside (57) C _ Apprend Desips Atatariel- Enphnwed Sanney Fonense

                                                                                                               )
 .s ~'

ANSTEC - ~

                  ..                APERTURE                           s e s c.3 , ,o       ,,,,,,,a g   LAns Alco Ava!?sh:2 on Aperturo Card (Note S)              M$)               (Note 2)   (Neta 3)   Vest MN                       (Note $7)
                                                                 .and Drain        ,
                          ,                                      .for         Justifir=#1a=

Vabe Position , , Actuator Actuation Type-A Type C for Not 5'mi=31 Safo ihmidewn Acddest .. Type Signal Type Test Test Testing Nemessential C O C P MSIS A.R,hk No No Note 7 Nonessential C C - SV Setpoint - Nonessential C C - SV Setpoint - Nonessential C C - SV Setpoint - Nonessential C C - SV Setpoint - Nonessential C C - SV Setpoint - Nonenential C C C S - Aeu Noneuential LI 14 Id E - RM C C HW - M Qonessential C C '

             .C                P, /          MSIS      A.R.M LO    O/C       HWf         .

M NonesseD% i. (JSIS A.R.M Essential -7 Nonessential '

O O/C P\ MSIS A.R.14 No No Noac 7 Nonessential O .

No O

O/C .

Matial

O O/C P MSIS A R,n( No No Note 7 Nonessential O O/C , P MSIS A.R.M Nonessential 0 O/C - - -

Nonessential 0 O/C - - - Nonessential

O O/C P MSIS A.R.M No No Note 7 Nonessential

O O/C P MSIS A.R.M Nonessential O O/C - - -

Nonessential

O O/C P MSIS A.R.M ho o Note 7 Nonessential

O O/C P MSIS A.R.M Nonessential 0 O/C - - -

Nonessential J C O/C E EFAS/AFAS A,R,M No No Note 7 Essential > C O/C - - - - C O/C E EFAS/AFAS A.R.M No No Note 7 Essential > C O/C - - - I W/228000/

w t:m 80+ Table 6.2.4-1 Containment Isolation Valve and Actuzioi bata (Coiii'd.) (Note 16) (Note I) Imtion Relative flow Direction to ' Relative to Valve Item Valve Egure No. Valve Contain. Contti -=# Arrangement Service No. No. Type ment (GDC) Nonnal 90 Personnel Aidock #2 - - Check Outside None - C Equahzation Line - Check inside (N/A) C 91 Containment Sump Pump - - Gate

Outside Out 14 O Discharge Line -

Gate

Inside O (56) 92 Containment Ventilation Units * - -

Gate

Outside Out Condensate Drain Header -

16 O Gate

Inside (56) 0 Check inside C 93 Reactor Drain Tank Gas Space to - -

Globe

Outside In/Out 25 0 GWMS -

Globe

Inside (56) O 94 Decontanunation Line -

Globe Outside In 21 If Globe Inside (56) If 95 Division i Hydrogen Recombiner - 6.2.5-1 Globe

Outside Out 23 Suction from Containment -

C Globe

Inside (56) C

           %      Division 2 Hydrogen Recombiner          -

6.2.5-1 Globe

Outside Out Suction from Containment -

23 C Globe

Inside (56) C 97 Division 1 Hydrogen Recombiner -

62.5-1 Globe

Outside in 4 Discharge to Containment -

C Check Inside C (56) 98 Division 2 Hydrogen Recombiner - 6.2.5-1 Globe

Outside In 4 Ducharge to Containment -

C Check Inside (56) C 99 Steam Generator Wet Layup - - Gate In Outside 9 14 Rectreulation Return to Steam - Check Inside Generator #1 (57) C 100 Steam Generator Wet Layup - - Gate Outside in 9 LC Recirculation Return to Steam - Check InsMe (57) C Generator 12 Approved Desips Material- Engheered Safety Fentures

 ./ '~'

l

ANSTEC - APERTURE

                                      ,v f}OQO                        Deskn ControlDocument Also Available On Aperture Card (Note 5)           (Note 4)        (Note 2)              Vent (Note 3)           (Note 6)                  (Note 17) and Drain for ,           Jusufication Valoe Position        Actuator       Actuaties            Type-A    TypeC      for Not       Essential /

fab Safe Shuhlews AN Type Signal Type Test Test Testing Nemessesdal C C - - - Yes Yes - N/A C C - - - AI O C P CIAS/HRAS A.R.M Yes Yes - Nonessential Al O C P CIAS/HRAS A.R.M AI O C E CIAS/HRAS Al A.R.M Yes Yes - Nonessential O C E CIAS/HRAS A.R.M C C - - ~ , Al O C E CIAS Al A.R.M Yes Yes - Nonessential O C E CIAS A.R.M LC If HW - M Yes Yes - Nonessential If LC HW - M Al C O/C E CIAS Al A.R.M Yes Yes - Essential C O/C E CIAS A.R.M Al C O/C E Al , A R.M Yes Yes - Euential C O/C , E A.R.M Al C O/C E A.R.M Yes Yes - Essential Al C O/C E CIAS A.R.M 4- C O/C

                                - ( - qgp j-                     Yes       Yes          -

Essentia!

O/C If HW M No No Note 7 O/C C - -

Nonessenaial O/C If HW - M No No Note 7 Nonessential O/C C - - - sm 9 Y/228000l-UL '*"%

4 a i System 80+ Deslan canarolDocument l l a 4 5 i Table 6.2.5-8 Hydrogen Mitigation System Igniter Locations ! Elevation Asimiuth Radius l i Tag No. I -tian (R.& Is.) (Degrees) (R. & Is.) j 1A* Reactor Cavity ICI Area 70+0 22 13-0 l l 1B Reactor Cavity ICI Area 70+0 314 13-1 l 2A' IRWST Area 87+9 44 44-11 2B' IRWST Area 87+9 316 44-11 1 3A* IRWST Area 87 +9 224 44-11 3 3B' IRWST Area 87 +9 136 44-11 l l ' j 4A MAVEC Area 89+6 355 28-2 4B* MAVEC Area 89 +6 351 35-8 5A* MAVEC Vent Outlet 100+0 16 43-7 5B MAVEC Vent Outlet 100+0 342 44-1

!           6A             ' E1.91+9 HVT Area                            99+0           154                     42-6 j
!          6B*               El.91+9 HVT Area                            99+0           206                     42-6

~ 7A* El.91+9 SG-2 Wing Wall 100+0 58 43-2 i 7B El.91+9 SG-2 Wing Wall 100+0 76 494 1 8B* El. 91+9 SG-2 Wing Wall 100+0 122 43-2 9A* El. 91+9 SG 1 Wing Wall 100+0 238 43-2 9B El. 91+9 SG 1 Wing Wall 100+0 256 49-0 10A El. 91+9 SG 1 Wing Wall 100+0 284 49-0 10B' 1 100+0 302 43-2 f El. 91+9 1 I

~./

g( q qJ+9 SG 1. 100+O IO4 '19~ O ., l I 1

      .%..;      Doelpr Ateewd- W Senoty Fenauen                                                                  Pope 6.2-232

i, k System 80 + Oesian control Document 1 i i

,         The emergency zone volume is 67.300 cubic feet. as shown in Table ISA-10. The maximum unfiltered j          rate into the control room emergency zone under accident conditions is 10 cfm. as identified in Tables 3          9.4-3 and 15A-10.

6.4.2.2 System Operation l j The Control Complex Ventilation System will be furnished with dual air inlet structures, structures 'A' and 'B' which meet the intent of paragraph C of Regulatory Guide 1.78. Each air inlet structure will be furnished with redundant Seismic Category 1. Class 1E radiation monitoring devices. Upon detection of 4 radiation levels greater than the allowable limits of 10 CFR 20 at the intake or upon receipt of a Safety

Injection Actuation Signal (SIAS), component control logic will automatically divert the control room j intake and the recirculation flow via the designated control room filtration unit. Upon failure of the .

3 designated filtration unit to start, the redundant filtration unit will start automatically. Additionally, j automatic selection and control logic is provided which compares radiation levels at both intakes and ] automatically isolates the more contammated intake. Section 9.4.1.5 presents a description of this i automatic selection and controllogic. This logic is active throughout the duration of the event, ensuring j that the dose for control room operator, evaluated for an accident duration of 30 days is both within the { guidelines of GDC 19 and As Low As Reasonably Achievable (ALARA). i I In order to safe-guard the control room from hazardous chemical relea ! j Class IE chemical instrumentation will also be provided. If structure 'A' detects a certain level of chemical pollution in the air, the instruments signal the dampers at structures 'A' to automatically close. ] As the chemical pollution in the air diminishes to a safe level, the instruments then provide a signal in . i l the control room which permits the operator to re-open the dampers. The same procedure would hold j true for structure 'B'. If the air at both intake structures becomes cb:anically polluted, the mstruments l at the intake structures would signal dampers at both locations to close, j .J The following passive safety features are available in System 80+ design to cope with the closure of both l intakes due to high concentration of toxic gas at both intakes. I } In case of closure of both intakes due to high level of concentration of toxic gas at both intakes, the

operator will shut down the control room air conditioning and filtration systems. This will eliminate leakage caused by the pressure differentials created by the fans. The control room temperature will be l

j allowed to rise to pressurize the control room. The operators will don themselves with face masks connected to a storage reservoir before the control room temperature rise exceeds 10*F. The air released from the face masks will travel away from the occupants as it mixes with the control room environment , i and create slight pressurtzation to mirunuze inleakage to the habitability zone. The face masks are at j locations accessible to the operators. The storage reservoir provides six hour supply of breathing air. After the operators start using the breathmg air, the air conditioning units will be started to keep the l control room temperature from rising. (( Detailed operating procedure to cope with the toxic gas accident ! will be prepared by the COL applicant. The exact location and detailed design of the storage reservoir

will be provided by the COL applicant.

] The COL applicant will develop an operating procedure to cope with closure of both intakes due to high ] concentration of smoke at both intakes.))l a 8 COL information item: see DCD Introduction Section 3.2.

             -           ~.- a.~                                                                              ,          em

i i System 80+ oesian controlDownsent Table 6.41 Control Room Doses from Design Basis Accident Integrated Dose (Real Control Room Scenasio Whole Thyroid Body Beta Gamma Skin 30 5 30 SRP 6.4 I imits 0.30 0.002 0.11

G1S 0.75 0.021 0.60
Failed fuel

2. Control Element Assembly Ejection (failed fuel) 5.37 0.21 3.83

Via SG Releases 24.0 0.025 0.37
Via containment

3. letdown Line Break Outside Cont.

r/

GIS < 2.27 < 0.015 <1.03

4. SGTR w/o LOOP 10.51 0.043 3.81

PIS

      'd 1.95                 0.043               3.79 o GIS v                '4- SGTR with LOOP i                                                                       9.52                  0.043               3.47

P!S ,

0.043 3.45 (

GIS re- SGTR with LOOP & Single Failure 1.80 A

PIS 26.75 0.32 25.3 4
GIS 30 0.32 25.3 3

, 4- 5. Fuel Handling Accident

In Cane *m= war 1.68 0.051 0.71 l
In Fuel Bldg. 1.68 0.051 0.71

6. Steam Line Break

SLBFPD-Failed Fuel < 22.61 < 0.0436 <0.80 l
SLBZPLOPD

                                  . pts                                       < 13.52                <0.007            < 0.22

GIS < 30 <0.022 <0.44

7. Ims of Coolant 30 3.88W 27.05 j

8. Feed Water Line Break

                               . MDNBR-Failed Fuct                            <781                  <0.015             <0.28

Over Pressure Casem

                                   . pts                                       < 3.98               <0.0021            <0.064-
                                   . G1s                                       < 8.90               <0.0064              <0.13

The gamma contribution due to shine through the control room walls and penetrations, from external sources such as contamment shine, cloud shine, skyshine, control room filters etc. is estimated to be 2.0 Rem and is included.
7tc control room doses resulting from the over pressure feed line break accident are bounded by the Steam Line Break (SLBZPLOPD) accident, am oenen neonannt . Engmeered Sonny Feeanne !*9* 6 "*

j "_ _ l System 80+ Denian Control Document i l 1 l 3. In the routing of CSS Class IE circuits and location of equipment served by these Class

1E circuits, consideration is given to their exposure to potential hazards such as l postulated ruptures of piping, flammable material, flooding, and non-flame retardant

. wiring. Adequate separation or protective measures shall be provided. I

4. Failures of nonsafety grade systems will not compromise the redundancy of the CSS.
I

Independence q

1. The environmental control system provided for each mdependent CSS train is powered j by the same emergency power associated with that train.

j 2. Power connections for CSS components are from two iManaMant electrical buses. ! 3. Four iMaaaMam vital instrument power sources are f.ovided for the CSS instrumentation, ] j e Monitoring 1 I 1. Provisions are made for the detection, containmem, and isolation of the mazinem q expected leakage from a moderate energy pipe rupture in one train. i j 2. Process instanmemanan is available to the operator in the control room to assist in assessing post-LOCA conditions. Instrinnantarian is listed in Table 6.5-2. l

Inspection and Testing

,i Inspection and Testmg requirements for the CSS are containeri in Chapter 16. Prior to initial i plant startup, CSS flow tests which comply with Section 6.5.4 are performed.

Chammtry and Sampling l 1. The CSS is designed for the followmg fluid conditions:

! Basic Fluid Water with 2.5 wt% H 3BO3

Long-Term pH > 7.0.

i 2. The Sampiing System provides a means of @iaia- remote liquid samples from the CSS 1 for chemical and radiochemical laboratory analysis. j

Materials

1. CSS piping and fittings are Seismic Category'I.

i

!                                  2.         Design and fabrication of the CSS piping and fittings conform to ASME Boiler and j                                             Pressure Vessel Code (B&PV) Section III, Class 2.

i i 3. Pipes and all parts in contact with the system fluid are austenitic stainless steel. 4 Amareweg Design Adesmet . Engmuserew Sedsfy Feemase page 6.E 4 1 C_ .__ . _ . . . - __

i System 80+ Deslan Control Document locking provisions are shown on the piping and instrumentation diagrams, Figures 6.3.2-1 A, 6.3.2-1B, and 6.3.2-1C. Valves SI-104 and SI-105 are local-manual vaives located in the CS pump suction piping from the IRWST. The valves are closed to provide double valve isolation of the RCS pressure boundary in the event that the CS pumps are to be used for shutdown cooling. 6.5.2.2.3.3 Relief Valves Protection against overpressurization of components within the CSS is provided by conservative design of the system piping, appropriate valving between high-pressure sources and low pressure piping, and by relief valves. Relief valves are provided as required by applicable codes. 6.5.2.2.4 Spray Noah and Nozzles I The C5S, including the spray headers, consist of two independent, redundant trams. Each train provides spray to three regions (I, II, and III) of containment as shown in Figure 6.5-1. The spray nozzles are

     =nached to and become part of the spray headers located in the upper portion of the containment building and below the operating floor. The nozzles disperse the spray solution throughout contamment in the form of droplets to increase the heat transfer surface. The spray nozzles and headers are oriented to provide maximum effective coverage of the contamment volume (See Section 6.5.3.1 for more                                    )

1 information of volume coverage). The spray nozzles are of the nonclogging type, designed to pass l particles having a maximum diameter of 5/16th of an inch. % The CSS uses Lechler (SPRACO Company) 1713A no or equivale . The nozzles provide a drop # size distribution which has been established by testing C . ue suitable for the fission product removal function. The CSS provides a nozzle pressure differential of 40 psid which fixes the drop size distribution. The mass mean drop size produced at this differential pressure is conservatively assumed to be 1000 microns. 6.5.2.2.4.1 AC-_- " of Spray Headers and Nozzles l The CSS headers and nozzles are arranged to provide maximum coverage of the contamment volume. The following information is provided for train A with train B information included in parentheses. Where no information is included for train B, values are the same as train A. All nozzles are oriented to spray downward at a 60 degree angle from the vertical unless otherwise noted. Reference Figure 6.5-1 for sprayed regions and location of spray headers. Region 1 is sprayed by 6 ring headers located 3 ft from the containment wall and a minimum of 30 ft 2 in, above the operating floor. Each train in Region I contains 126 nozzles. Ring 1 is located at a radius of 14 ft 8 in. (15 ft 1 in.) from the centerline of contamment at an elevation of 252 ft 11 in. (252 ft 10 in.). The ring has 6 nozzles, all of which spray toward the center of containment. The nozzles are equally spaced on the header with alternate nozzles oriented to spray at a 15 degree angle from the vertical. Ring 2 is located at a radius of 26 ft 2 in. (26 ft 7 in.) from the centerline of contamment at an elevation of 250 ft 5 in. (250 ft 4 in.). This ring contains 9 equally spaced nozzles which are oriented to spray toward the center of containment. Ammat Design aMwM - Enskment Sekty Feeaner Pope 6.510

1 System 80 + Deslan controt Document l , j ! The sprayed regions are described more fully as follows: j j

Region I

The total free volume of Region I is 1.328 x 17 ft3 and includes the following:

1 i 1. Volume above operating floor level at elevation 146 ft 0 in. and inside the crane wall. I j 2. Volnma i-ida 'ha =ragenerator cavities, j 3. Volume between elevations 91 ft 9 in. and 146 ft 0 in, which is sprayed throug J ft x 10 ft grated openmg at elevations 146 ft 0 in. and 115 ft 6 in. ] '

4. Refueling canal (portions of which are not obstructed by equipment located above) i l S. Core Support Barrel Laydown Area. l

] 6. Upper Guide Strucmre Laydown Area. j 4 i e Region Il j The total free volume of Region 11 is 1.317 x 1& ft'. His volume is located outside the crane ' wall and above elevation 115 ft 6 in. and includes the portion of the refueling canal which j

extends outside the crane wall. I a l 4

e Region m l I I j The total free volume of Region m is 0.077 x 11 ft'. This volume is located between elevations i 115 ft 6 in, and 142 ft 0 in. and radii of 55 ft and 65 ft frou the centerline of corummment. ) In addition to the corwsmment sprayed regions there is an unsprayed region of the contamment identified 2 as Region IV. Region IV has a total free volume of 0.615 x 1& ft and includes compartments such as j j the following:

1. Cavity around and below reactor vessel.

l i l l 2. Pressurtzer cavity. i 4 i i i 3. Heat ~~haa-er compartments. t i j 4. In-core chase. ' Annular space above the crane wall. 5. The breakdown of the contamment volume into sprayed and unstrayed regions is summanzed in Table 2 j 6.5-4. The total sprayed volume from all sprayed regions is 2.722 x 17 ft with a total corummment free i volume of 3.337 x 17 ft 2This volume represents 82% of the contamment free volume. He remaming i 18% is assumed to be unsprayed. t l 1 ! Anweved Deeps A0eenrod Eneheered Seenty Feeeeen Peee 6.6-16 i

i i i System 80+ Deskn Control Document The addition of trisodium phosphate dodecahydrate to the containment spray solution results in the following composition in the IRWST:

e 4400 ppm boron as boric acid 1

e 2000 ppm PO. 6.5.3.3 Airborne Fission Product Removal Coemcient The removal of airborne fission products in particulate form is considered to a first order with respect ! j to particulate concentration and is mnrh,matically described as follows: b e -AC A / (6.5-1) dt p , J O 'J > I which integrates to gA Y c v D C = C,e u / 4 (6.5-2) where C = particulate concentration in the contamment atmosphere A = particulate removal coemcient C, = initial particulate concentration in the contamment atmosphere i i t = time after the spray operation The particulate removal coemcient is related to the spray parameter as follows: 3 I A=e (6.5-3) 2Vd where e = removal emciency h = average spray drop fall height F = spray flow rate V = containment sprayed volume d =- drop diameter Anuwwod cemen unserw Enpmered Sekty Footwee Pope 6.519

System 80+ Oeelan ControlDocument ' i j condensation has not been included in the present analysis, although the impact of having included this j effect would have been significant as discussed in Section 15.6.5.5. l One of the inputs to SWNAUA is the particle size distribution of the aerosol source. For the System j 80+ DBA LOCA analysis these distributions are lognormal with the following geometric radii (r,) and i standard deviation (a): i l j r, = 0.075 gm. a = 1.56 for the early release of gap activity, and

     ~~"7 j                r,         =      0.4 gm, a = 1.46 for the release of activity associated with fuel melt.

i b \ 4 These distribu@ns are based on RAFT predictions for STEP-1 (see Reference 4). -i j Another of the inputs to SWNAUA are the particle densities. For the System 80+ DBA LOCA particle densities are based on SASCHA expenmental results from Reference 5. "Ihe values used 3. l gm/cc for the gap release and 4.6 gm/cc for the melted fuel release. In applying these densities and size distribmions to the SWNAUA model for System 80+, no credit has been taken for condensation of water on the particles; this is in addition to having neglected CsOH hygroscopicity as dise=M above and 4

 '                                                                                                                                I further below.

In order to calculate the effect of steam condensarion on the spray droplets, the total amount of water removed from the conenmment . simp.m as a function of time in the thermal-hydraulic analysis is

apportioned according to heat removal by the strucares and the sprays. Only condensation on the sprays
is included in the fission product aerosol removal calculation of SWNAUA. "Ihe conservatism of having made this assumption is discussed further below.

Ir. deternumng the effectiveness of the spray (i.e., in danaiag the spray " lambdas") only the effects of the spray have been considered. Diffusiophoretic deposition of aerotols on the containment structural heat sinks and sedunentation in the sprayed region have been neglected. There is considerable ' conservatism in having done so, parucularly in the in nadiate post-blowdown and core quench phases in < which steam condensation rates are the highest. Another unportant phenomenon which has been neglected, as discussed above, is the hygroscopic treatment of certam fission product aerosols; e.g., , CsOH. CsOH is extremely hygroscopic, and the effect of having neglected the hygroscopicity of CsOH (which makes up about 25% of the aerosol mass released to conramment) is to underestimate the particulate size distribution which, in turn, leads to a low estimate of spray effectiveness. Not all of the airborne fission products are in particulate form; iodine will also appear as 1, 2 HI. and organic iodidas in the cone ==mant atmosphere.2 I and HI are reactive and will tend to plate out on surfaces. '!he major fraction of available surface is the sepa* aerosol; therefore, the same spray lambda is applied to non-organic gaseous iodine in the sprayed volume as to particulates. This is a conservative assumpoon because the non-organic gaseous iodine spray removal lambdas would otherwise tend to be somewhat greater. I Consistent with the assumption that non-organic gaseous iodine is adsorbed upon the particulate, there has been no maximum DF applied to its removal. In fact, since organic iodme is assumed to be 0.25 % of the iodine released to containment (see Table 15.6.5-2), and since no credit is taken for spray removal , of organic iodine in contamment, a residual quantity of airborne elemental iodine eq J to or less than 0.025% of the total iodine released would be negligible (i.e., < 10 percent of the organic iodine contribution) in te .ns of impact on dose. With an assumed initial percentage of 4.75 %, a DF of at least 190 would be needed to achieve the " negligible" percentage of 0.025% for elemental iodine. Annreved Donen neesend . Enoneemt Senoty Feenwee Peee 6.6 2r l

System 80+ Design ControlDocument i In order to calculate the effect of steam condensation on the spray droplets, the total amount of water removed from the containment atmosphere as a function of time in the thermal-hydraulic analysis is apportianed according to heat removal by the structures and the sprays. Only condensation on the sprays l is included in the fission product aerosol removal calculation of SWNAUA. The conservatism of having made this assumption is discussed further below. In determining the effectiveness of the spray (i.e., in defining the spray " lambdas") only the effects of the spray ha,e been considered. Diffusiophoretic deposition of aerosols on the containment structural heat sinks and sedimentation in the sprayed region have been neglected. There is considerable conservatism in having done so, particularly in the immediate post-blowdown and core quench phases in [ which steam condensation rates are the highest. Another important phenomenon which4has neglected, bee # as discussed above, is the hygroscopic treatment of certain fission product aerosols; e.g., CsOH. CsOH is extremely hygroscopic, and the effect of having neglected the hygroscopicity of CsOH (which makes up about 25% of the acrosol mass released to cocainment) is to underestimate the particulate size distribution which, in turn, leads to a low estimate of spray effectiveness. Not all of the airborne fission products are in particulate form; iodine will also appear as 21, HI, and organic iodides in the containment atmosphere. 12 and HI are reactive and will tend to plate out on surfaces. The major fraction of available surface is the suspended aerosol; therefore, the same spray lambda is applied to non-organic gaseous iodine in the sprayed volume as to particulates. This is a conservative assumption because the non-organic gaseous iodine spray removal lambdas would otherwise v tend to be somewhat greater. Consistent with the assumption that non-organic gaseous iodine is(femoved as depositedupofarticulate, there has been no maximum DF applied to its removal. In fact, since organic iodine is assumed to be 0.25% of the iodine released to containment (see Table 15.6.5-2), and since no credit is taken for spray removal of organic iodine in containment, a residual quantity oglemental iodiiihairbomelequal to or less than 0.025% of the total iodine released would be negligible (i.e., <10 percent of the organic iodine contribution) in terms of impact on dose. With an assumed initial percentage of 4.75%, a DF of at least 190 would be nemied to achieve the " negligible" percentage of 0.025% for elemental iodine. srn$x m With a containment free volume of 3.34 X 106 ft3and approximately 7@10 5gallons of primary coolant and IRWST inventory available, the partition coefficient necessary to obtain the DF of 190 would be approximately 6800. Such a partition coefficient would be readily obtainable with a spray pH of 7, or even slightly less. In Reference 6, for example, a pH of 7 limited conversion of I to 12 to only 0.03% even after 24 hours exposure to 0.35 Mrad /hr of radiation (about 40% greater than a or System 804 IRWST). To get a partition coefficient less than 6800, the copversion of I to 1 2 wou have to be much greater than 0.03f,Te.n. 0.08% at a temperature of 212_*tgTherefore, a DF of 190 i conservative for a pH of 7. Y -m M% fw? -b

W- e A .La 9 1p W W Jtx.u wa.=>. g

                                                                                      -k. Q k The pH of 7 is expected to be reached in approximately 2.5 hours after the LOCA (secTection 6.5.3.2).

Nevertheless, it is reasonable to consider (1) what the impact on the dose assessment would be if the rate of pH increase were significantly slower than that yielding a pH of 7 at 2.5 hours, and (2) what the long-term impact would be of lowering pH due to acid formation. These two concerns are addressed in the following sensitivity study. Approved Design Material- Engineered Safety Feenwes Page 6.5-22

System 80+ Design ControlDocument l 6.5.4.2 Inservice Inspection and Testing l The CSS is designed and installed to permit inservice inspection and testing in accordance with ASME Boiler and Pressure Vessel Code, Section XI. The complete schedule for CSS tests and inspections is described in Chapter 16. 6.5.5 Emergency Containment Spray Backup System i The Emergency Contamment Spray Backup System (ECSBS) is used as an alternate means of providing containment sprr, w .he event of a beyond design basis accident in which both CS pumps, both SCS pumps, and/or th: V#ST are unavailtsble. The ECSBS pumping device operates independently of the site normal and energency AC power sources and is capable of being aligned to the spray header of either CSS division; see Figures 6.3.2-1A and 6.3.2-1B. l l The ECSBS design includes the following features:

1) an eigit1'mh diamet tee connection to the IRWST recirculation line of each CS pump;

                    'l l

2) an extension of the eight inch diameter Safety Class 2 piping from the tee to the exterior of the Nu:: lear Annex;

3) external connections, at or near grade level, for the temporary hookup of an external source of water; j i

4) a portable pumping device (e.g., fire truck) that has a minimum rated flow of 750 gpm;

5) all necessary hoses, 6ttings, and spool pieces, which would be stored with the pumping device or in the proximity of the piping connection to the IRWST recirculation line.

The ECSBS is assumed to be placed in service 24 hours after a severe accident to prevent a catastrophic ! failure of the containment. The ECSBS flow rate provides sufficient heat removal to prevent containment pressure from exceeding ASME Code Section III Service Level C limits. The design take! i nto account expected radiation levels and shielding requirements for any required local operator actions; e ECSBS (f( 4 pumpingevice is beyond the scope of Design Certification and shall be discussed in the site-specific g._6 SAR.h The ECSBS pumping device;and the connection to IRWST recirculation line tee will also provide an alternate means of introducing water into the containment, via the IRWST, after a severe accident. By repositioning remote-manual yalves in the IRWST recirculation line of either CS pump, the ECSBS equipment and external water source can be used to refill the IRWST to provide long term flooding capability in the event of a severe accident. Refennces for Section 6.5

1.

Licensing Design Basis Source Term Update for the Evolutionary Advanced Light Water Reactor," Advanced Reactor Severe Accident Program (ARSAP) Source Term Expert Group, September,1990. I g yt.- QWWtdbyg .

         ~
                                     -  3>-

i Agowd Dukn MtwM Enn: SMuty Fuaan Pege 6.6-25

    =_                                                                _                                                            i System 80+                                                                       Deslan Control Document I

2. "SWNAUA VER02. LEV 00 - Aerosol Behavior in a Condensing Atmosphere - Diffusiophoresis and Spray Version on a PC." Stone and Webster Engineering Corporation, NU-185, M:.y.1993 (SWEC Proprietary).

m W., "NAUA Mod 4: A Code for Calculating Aerosol Behavior in

3. MH., Koyro, Schock LWR Core Melt Accidents, Code Description and User's Manual. Preliminary Description,"

l l Laboratorium fur Aerosolphysik and Filtertechnik I, Projekt Nukleare Sicherheit, ! Kernforschungszenmun Karlsruhe, March,1982.

4. Im. et. al., " RAFT: A Computer Model for Formation and Transport of Fission Product Aerosols in LWR Pnmary Systems " ANS Topical Meeting, " Fission Product Behavior and Source Term Research," Snowbird Utah, July 1984. ~

i

5. Albrecht and Wild. " Review of the Main Results of the SASCHA Program on Fission Product Release Under Core Melting Conditions," ANS Topical Meeting on Finion Product Behavior and Source Term Research, Snowbird. Utah, July 1984 l

6. Beahm, E.C., C.F. Weber, and T.S. Kress, "Iodme Chemical Forms in LWR Severe Accidents,"

NUREG/CR-5732,ORNL/TM-11861 July 1991, l N l l p, g, y,,, W W M

W SektvFoemeen

f 1 8 1 l i , ) ' i System 80+ Desian contreu Document l i i Table 6.5-2 Containment Spray System Display Instrumentation i i i identification No.

;                                                                                                                                                 i.

Parameter Train 1 Train 2 Type of Readout Indicator Location l J CS Pump Discharge P-338 P 348 Indicator / Alarm Control Room j Pressure ! } CS Pump Suction Pressure P-310 P 320 Indicator /Alarmm Control Room 4 l Pump FloWu F-338 F 348 Indicator Control Room . I i F-302 F-305 Imhcator/ Alarm

i i Sprty Header Isolation SI472 SI471 Open/ Closed Status Control Room i Valve Position 0100% Position Control Room

!                                                  ,                                      Indicanon i

j Spray Header Block Valve 51487 S1495 Open/ Closed Starus Control Room !

Position 1

j IRWST Return Line SI486 51496 Open/ Closed Stams Comrol Room j Isolation Valve Position d IRWST Return Line Flow SI457 S1458 Open/ Closed M Control Room Control Valve Positio : fl00% Position saae= nan Control Room , l CS Pump Discharge Manual SI-488 SI-489 Open/ Closed Status Control Room . , i j Isolation Valve Position Pmnp Stams* CSP 1 CSP 2 tade=nar Lights Control Room } Pump Motor Current I-338 I348 Indemar/ Alarm Control Room

CS Heat Exchanger Outlet T-303X T 303Y Indemar Control Room Temperature l

i l

                                                                                                               .lv      p                      ,

I

                                                                                        ^            '

y , mYp ' Kv

                                                                             / i l
                                                                                    +J /

i 4 l 8 Required for post-accident monitoring. Specified function provided when CS pump is used for shutdown cooling. l i Anweved Deep, neesener anoneeered senny foeewee rose s.5 2s i

s _- t f i i System 80+ oesten canaat occumenf i' k 1 Table 6.5-3 Containment Spray System Failure Modes and Effects Analysis s.- se- ran.e M.de Ca.e Sympi.- ami wein.d .f nahm.= n==w tacal ENects n e.,e== Campensasing and Other 1 laduding Provansa ENecte j

;                                                                             Dependsat rea==

3 Parauel Valves are j 1) Cop'mianaent Spray al Fads closed Corrosion. tracha= ara' ENecove loss of Law flow inadsag. operamt error one consmument adacemon F- redundant normany locked j Peep Succon

'                                                                             spray pwnp              338.               cons =M         open isolaeon Yahe F-348; penodic spray pee i                    SI-106 51105                                                                             issoas Mo go,,k /,

{ Noas requuod b) Fans opes Corrossa No effect dunag Penodic tesa CSS operamon J Consavanans Spray Mm.hancal tadure. ENacave loss of 14w flow Pasauel k 2) a) Fads ao redundant wf' Pump 1. 2 peep on eiscencal fanuse one comaumasat indicanoa, j spray pedi F-338. t - -- 4 CSAS 348;penadac speny pedi

                                                                                                      ""'8                                                              l ESecove loss of         Low pressure       Parnuel        Valves are j                3)  C-- Spray           a) Fads        Corrossoa.

r==h===e=' todos, one commensat adacamos redundaar nonnauy locked Peep Discharge closed spray paa P-338. e- - open

Isolanom Valve operaer error P-348; valve syssy pee 1

51488. SI489, pommes pae adicaser. ! penodec .-- ! / b) Fans open as , No effect on Penedic saames Nousrequued 4*dc/ j ( at H safuey opesumos g Conauessat Spray a) Fem 8 Elect. Malf. Poemhis isolanos Compenson Raheadant assa 4) j Pump Flow adscanna of aneceedCS widi sedundmet /jg* ladscamon F-338. inua escanor,wie j X F-39 di **= psecess monumsat- l f egna and valve l l I Poemas 1 a ncemons 9

5) Commaamses Spesy a) 14ss of InnsfRcasa ev Diamanhed abday High Parausi Heat Eschanger 1. cooing coohng want flow, of subsysmato esaperesse fedeadas ea===-

2 essesesve foules psoveds igdsceans from 4 sempernaus. & ( ) spray pee psessus T-303X. l suppseemon T 303Y 4-wiesa es cammmensat 4 dunas ) ince ina.a made of operanos 4

l i

I i i 1 1 Angrevent Deenpn neeennaf Enganeereef SeMy Feeanne Pope 6.510 ] i s

L O" System 80+ Desien Control Docannent Table 6.5-3 Containment Spray System Failure Modes and EfYeCts Analysis (Cont'd.) No. Name Fadure Came Syuspteens and Method of Detectase laberent Remarks and Other Effects Mode Local Effects Campanesung lachading Provunna Depsedest Falksres

6) Cori=innunt a) Fads Corrosson. Effecove loss of Valve posioon Parallel Valves are normally open Spray Header closed mechansal one conaunment indsanor: penods redundant Block Valve bandang. spray path usang conamment 51487 operaer sprey pse 51495 enor electncal *,

fadure b) Fans Mach ==<al No effect IsunnsPenods usang Nons requuod open fadum CSS opersoon

7) Coneunment a) Fads Corrosson. Effecove loss of Valve posioon Parallel Spray Header closed inachaaical one consannasat undsator, penods redundant Iseignon Valve bedang. spesy pash tasang cosasament

     <:     (                             elecincal                                              spray paa
              @ S1472                     fadme 4,e' i#                            b) Fans      Mechancal    No effect dunas Valve poemon              Nome required    Valve u nonnauy locked open         fadas.       CSS operance     indmemmr. penods                          closed elecincal                     asmes fasime                   f""""A                  A B)   Conaanment     a) Nozzle    Foressa      No commaa-asses  Low flow andmanaf Pamuel                  The nozzles have an appions-Spray Nozzles  blockage     obpecs a     spray flow no    F-338: F 343:            redundastC5      meesty 3/8* spray onAce and
                                          -a= ant      affecsed nozzle  penods usang             essa             wel not be subject to cloggag spray hass,                                                             by paracles less than 5/16* in cormsson                                                                ======= danseasson The sprey soluson is compiseely schie and solubie at all temper.

ammes of innersst a the e-a== ant and therefore wdl not precapuses or otherwise teserfere with nozzle perfor-ranace

9) Coamin=== a) Loss of insufhcesa Possibly damage Penods assung Parallel Spray Muuflow coohng ca-ra=== to one CSP redundant Heat Exchanger coohng ca== -

t. 2 woest flow, spesy tram emessesve feelms

10) Coa ====== a) Fads Conesena. Diversion of Valve posioon Puauel Spray IRWTT open mechmascal flow from C3 umhcanor, penods rad =wa=

bedes, Resurn Las header tesung Flow Consol operesor spesy nam. Valves enor serws isolanon 51457 eiscencal valve 51458 fadme b) Fans Corrosion. No effect dunas Penods assung Nons required closed rnschansal CSS opersoon beding Amweves Demon Aseannet esmeerar senery poenwee repe s.s Jf

System 80 + Deslan C ntrol O cument Table 6.5-3 Containment Spray System Failure Modes and Effects Analysis (Cont'd.) Ca.e sr., .ms aan u.ined.fpoi ins immrom Re ru aan ouer Efr.cis i s.. same ranure Mede Lacal Effects C-r====g faciudag Provisies Dependent Fautwee , a) Fads Corrosen. Diversen of Valve posson Parallel 111 Contamment Spray IRWST open mechanscal flow from RCS indmasor. penods re-ta= header wscas e - ==== 2 Rentm Lme bedang. ! operaer sprey tram. 1solanon Valves ermr. senes isolaeon

                     $1486.

elecencal valve 51496 faites No effect dunas Penods sesang None requuod b) Fans Conosen. closed enschansal CSS opersoon bedes

12) IRWST Isolanon a) Fads Corrosson. Effecove loss of Low flow indsanon Parallel i==chaaral one conmuunent F 338.p48: penods reduadant

          - A        k             -   closed buuhas,       spray puenp             tesang '               e-S1304 51 305                       oper c -                                        (, p .       sysay pedi ersor.

/

       %      (      '

gjecWeal fedme Corvosen No effect on Penods sesong Noen sogared b) Fans opes CSS opstamon

13) CS PuagwSCS a) Fads Conosson. Effecove loss of Law flow =temaam SCS is Pump Succon open raschenecal one comeus- F 338. nonsauy Cross <onnect bedsas, most spray F-348; hsgh CS peep esoissed Valves operaser peep sucaos psessus ,

51-340. error, umhcanon F 310. SI-342 elecincal P 320: penods fadme sesang; valve posmon umhcasor b) Fads Conossoa. No effect on Peimds usang; None seqamed closed machescal CSS opersoon valve posmon buuhas. umhcasor i -t/., elecencal

        "     N,                                    fadme

14) CS Puenp/SCS a) Fads Conomen. Effscove loss of High flow uuhcanos SCS is Pu harge open seechanecal one conam- F-338. noneauy Cross <onnect beems. meat spray F-348:ynods neolased Valves opener pump sesmaggalve posioon SI-341, enor, edscason SI-343 elecencal a) Fads Meck. None Penods tesung Shad flange Valve is nonaally locked 15)

Valve open banhas provides closed at valve isonsoon

                       $1-454 b) Fads     Mech.          None                  Penods resung closed      buutag I.

p A

                     /a) A                                                                                    '
                   ,u         9                .
                        .s V ,fyW, 4preveef Doesper Afesorest. ErspriseroW Soflerr Fooewee                                                                                Pope 6.5 J2

f' TABLE 6.5-3 (Cant'd) (sheet & of 6) CIElf AIISUT SPRAT STSIDI I AllLSE IgBE$ AaB &FflCl$ All4LT$1$ 3rdweent Bemerks arut

                                                                                                                                             $wema eruf Local Ef fects                                                                                                                                                             Commermatitu Provisfort felture llede                                               Cause               Itsidfu peerufernt Fellwes                                                                                              IIsthed of DetectIan                                                                   other tffects se a            name al Felts                                Cerrosion,                              E f f ective loss of one                                                                                                 Low flow irufication                                         SCS is norsully

13) C5 Pum/SCS contairunent sprey pump F-338 F-348; high Isolated Psmp suction open sechanical binding, C5 pap suction Cross-Comect pressure trufication valves operator error, electrical P-310. P-320, 51-340, 58-342 periodic testing, fatture valve position Indicator Corrosion, No effact en C$$ operetion Periodic testing, Mone re pired b) Fails vatwe positien ctosed mechantcet binding, indicator electrical falture JCS is normally

14) CS Ptsp/5CS a) foits Corrosion. Effective toss of one Nigh flow anchenical conteiruunnt sprey ptmp indication isolated Pup Discharge open F-334, F-344; Cross Comect biruffne. !,

operater error, perledic testing, valves vetve position

               $1,341, 51-343                                                                           electrical                                                                                                                                                                                                                                  _.

fatture trusica

                                                                                                                                                                                                                                                                             ..                         ..t.or              -- -

noeffec[onC$$ operation mone re wired Mh' b) f aits closed Corrosion, mechanical biruding, Periodic testens, valve position indicator i

s electricot failure

                                                                                                                                                                                              '~
                                                                                                                                                                                       ~-~'                                                                                PericdiC testing                                         Stitui flange          Valve is ruirmally a) Falle                                   Beech, birufing                      Mone

15) PCP$ isolation provides isolaticn toctted closed at valve open _.

vatwe hirufing mene Perlod6c testing 51-450, 51 454 b) FatLs leech ctosed i

                                                           /
            /C                   V h

4s,% )i j sf , i

              -                                              I                                                                                                                                                                                                                                                                   Amendment V ff                                -3                                                                                                                                                                                                                                                                      April 29, 1994

   -_- . - - - . - - -                        . . -.            -    -.         -=          -        - -      - - . _ - - - -

i e System 80+ Design ControlDocument l ! 6.6.2 Accessibility and Inspectability 3 - j Accessibility to equipment for nance, teTsefpection is a basic element of the System 80+ 1 design process, as indicated le 1.2-1 fitem A.2 ovisions have been made in the design and ! layout of Code Class 2 and 3 Systems to aW6w Torcompliance with the inservice inspection requirements i contained in ASME Code Section XI, Articles IWC-2000 and IWD-2000, and as defined in the

inservice inspection program. Code Class 2 and 3 components requiring inspection will be designed for
and be provided with access to enable the performance of Section XI inspections onsite. Systems and i components are designed such that design, materials, and geometry do not restrict inspections required by Section XI of the ASME Code.

I J Welds and other areas requiring periodic inspection are made accessible. Reinforcing pads, supports, i piping and equipment have been located not to obstruct welds. Insulating materials are removable to l provide accessibility for the required in-service inspection. l ! 6.6.3 Framination Techniques and Procedures 1 I The examination techniques to be used for inservice inspection will include radiographic, ultrasonic, magnetic particle, liquid penetrant, eddy current, and visual exammation methods. For all examinations, } both remote ard nunual, specific procedures will be prepared describing the equipment, inspection j technique, operator quali0 cations calibration standards, flaw evaluation, and records. These techniques and procedures will meet the requirements of Articles IWC-2000 and IWD-2000 in the ASME Code,

Section XI edition in effect as stated in Section 6.6.1. Pre-service Inspection (PSI) and subsequent: Inservice Inspection (ISI) will be conducted with equivalent equipment and techniques.

i ] For the preservice inspection, all of the items selected for inservice inspection shall be performed once { in accordance with the ASME Code Section XI, Subsections IWC-2200 and IWD-2200, with the j exception of the exammations specifically excluded by ASME Section XI from preservice requirements, j such as the VT-2 exammations for Category C-H. j For Class 2 and 3 piping systems subject to volumetric and surface exammation, the following piping j designs are not used: i ! 1. Valve to valve t

2. Valve to reducer l

1 3. Valve to tee _ l 1 4. Elbow to elbow ! t 1 1 5. Elbow to tee !

6. Nozzle to elbow

) 7. Reducer to elbow l J 8. Tee to tee i

 !      9.             Pump to valve i

Antowent Denip ntonerin! Enebenereef Sekty Feeener Pop 6.6-2 L _ - - _ - ,_.

Syntem 80+ Deslan Control Document

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Repluu 6q af R6h b eeve cq.

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Tnh F-geri Tokn 44a(l 3 Pl aoL l I . z _ a'L ) ( h Rgve l l 7 i

                                                                                  @lf l<%
                                                                                     $65Bb t.

4 l Reactor Building Plan View of Upper Dish Figure 6.8-1

    " %'9" MeterW EnM 5,4 y                                                    Page 6.812

l \ System 80+ Design ControlDocument e The test sequence described in IEEE 382-1980 or the actual service sequence, whichever has the most severe operating conditions, is used during operator qualification tests. j i e The valve operator is tested under the severest environmental conditions (T P, RH, Radiation) that simulate the conditions to which the valve operator is expected to be exposed during and following a DBA.

                    -o           The radiological source term for qualification tests is based on the same source term used in                 ,

Regulatory Guide 1.7 taking into consideration the containment size, beta and gamma radiation. 7.1.2.26 Confonnance to Regulatory Guide 1.97 The design of the post-accident monitoring instrumentation and information display via the DPS and DIAS is described in Sections 3.1 and 7.5. The design conforms to Regulatory Guide 1.97, ;

                      " Instrumentation for Light-Water-Cooled Nuclear Power Plants To Assess Plant and Environs Conditions                   l During and Following an Accident."                                                                                      '

i 7.1.2.27 Confonnance to Regulatory Guide 1.105

ll7he generation ofsafety system setpoints conforms to ISA-S67 %-1987, "Setpointsfor Nuclear Safety Related instrumentation Used in Nuclear Power Plants. "]i l The setpoint methodology is similar to that explained in CEN-278(V), " Selection of Trip Setpoint Values l for the Plant Protection System," submitted on the Palo Verde Nuclear Generating Station Unit 1 Docket, j STN-50-528 and approved by the NRC. The environment considered when determining errors is the - most detrimental realistic environment calculated or postulated to exist until the worst case time of the i required Reactor Trip or Engineered Safety Feature Actuation. This environment may be different for different events analyzed. For the setpoint calculation, the accident environment error calculation for process equipment uses the environmental conditions up to the longest required time of trip or actuation that results in the largest errors, thus providing additional conservatism to the resulting setpoints. ' The reference leg heating component uncertainties for steam generator level also take into account ' l pressure and temperature variation within the steam generator. For all temperature and pressure setpoints, the trip will be initiated at a point that is not at saturation for the equipment. For level setpoints, no analysis setpoint is within 5% of the ends of the level span. 7.1.2.28 Conformance to Regulatory Guide 1.106 Conformance to Regulatory Guide 1.106, " Thermal Overload Protection for Electric Motors on Motor-Operated Valves," is accomplished as follows. Thermal overload protection devices are not used in safety-related motor-operated valve control circuits. Thermal overload signals are used only for status annunciation. The ESF-CCS, as described in Section 7.3, has the design capability to provide MOV thermal overload status which is available via the DIAS and DPS described in Section 7.7.1.4 and 7.7.1.7. NRC[aff approval is required prior to implementing a change in this information; see DCD Introduction Section 3.5. Approveef Design nGatoriah knetrumentation amt Controer Page 7.116

l Pladr Pdech'" Sh (PPC avl Epm A 5% Feah> ret l cop O& syk ( E 5F- CCQ Syst:m 80+ \ ' Design Control Document A defense in depth approach is employed to eliminate common ode software errors as a concern for j the Nuplex 80+ instrumentation and control systems. This approa h is summarized as follows: f u Deterministic Desien - The algorithm execution in the Nuplex 80+jent:1 A p e cde cy::= d is deterministic. This means that all data is updated on a continuous cycle and all programs execute on a continuous basis, without interrupts. This approach makes the software easier to design, verify and validate. The potential for hidden errors is significantly lower than in other designs which include multi-tasking, event based execution, event based data communication, or interrupts. None of these I non-deterministic features exists in the Nuplex 80+ ee=~e' =A p ~"en syre , Pm .e E w -c.cs. c._ Field Proven Products - Operating system software for Nuplex 80+ I&C systems is selected with at least 3000 operating years and at least one calendar year of field experience in similar applications. These l products are mature and, therefore, judged to be free of infant design errors. Verification and Validation - For custom software generated by C-E, a comprehensive V&V program is employed, including independent document review and independent test. The V&V program for Nuplex 80+ is described in the Software Program Manual and its associated references. Independence is maintained between software development and verification personnel. Utility-Owner configuration controls are also imposed throughout the software life cycle. The V&V program minimizes the potential for introduction of common mode software errors during the design phase and during commissioned life of the system. l Seementation - Within all Nuplex 80+ systems, including the PPS ESF-CCS, and Process-CCS functions i are divided into separate processors. Segmentation within each PPS channel ensures that two different trip functions are available in two nparate processors for each design basis event. Similarly, within ESF-CCS Division A and B, ESFAS functions such as SIAS and EFAS are distributed to separate control processors. Within the Process-CCS critical plant control functions, such as inventory control, heat removal, etc., simultaneous errors in these multiple processors is minimized, since functional diversity is utilized and since software execution is asynchronous. l Diversirv - Diversity offers the final defense against common mode failures. All critical safety functions,

such as reactivity control, inventory control and heat removal, can be controlled by both the control l systems and the protection systems. These systems are functionally diverse, as are the fluid / mechanical l systems they control. In addition, to correspond with the hardware diversity of these fluid / mechanical systems, both hardware and software diversity is employed between control and protection I&C systems to eliminate the potential for common mode failures to affect both the control and protection functions.

This diversity exists in all software based aspects of these systems, including controllers, multiplexors, communication networks and MMI devices. This same diversity philosophy is applied between DIAS and DPS to ensure availability of control room information. Independent of the above design features, System 80-F implements a means for manual actuation of Engineered safety Feature functions using two safety grade channels which utilize hardwired inputs that bypass all data links, network communications, and all computers with large software applications. Switches located in the Main Control Room provide for system level actuation of two trains of safety injection and one train each of containment spray, emergency feedwater, closure of main steam isolation valves, closure of containment air purge valves and closure of a letdown isolation valve as shown in Figure 7.3-25. The switches for safety injection, containment spray and emergency feedwater have 3 positions, as follows: normal, actuate and stop. Approved Design Meterial-Instrumentation and Control Page 7.3-17 l

Srtem' 80 + Deslan controlDocument 7.7 Control Systems Not Required for Safety l 7.7.1 Description The control, instmmentation and monitoring systems whose functions are not essential for the safety of ; the plant include instrumentation and control equipment not addressed in Sections 7.2 through 7.6. l l Each of the Nuplex 80+ system functions have been analyzed to determine the characteristics it must I have to meet the required sections of regulatory guides and standards that apply to that system (Reg. Guide 1.97, NUREG-06%, NUREG-0737, IEEE-497, etc.) as well as desired reliability and performance characteristics. The following chart provides a definition of the key characteristics associated with the Nuplex 80+ indication and control systems-

                   #Acinbn%-beep Tyc i Non.1E/,Ca c vu i-("CS, ",xe..-CC         DPS)

No Credited Redundancy l Non-seismic (with exception of MCP/RSP mounted equipment) Joint Verification ; Joint Validation ! Multi-task CPUs

         -       Interrupt Driven'(" races-CCS, DF5) "
               - M" .u.T. !i;;ci, g. (,"C';) C~

Not Completely Deterministic On-line Changeability Non-restrictive Data Communications @rr CCS, DPS) - I imiled ".crise n m c~ ~ u.g;,;;g,,,,,;g,neg:ge,(nc3) - o ro 1 % h ati" TiPC E Non-1E/f.,%v.j !! (DIAS-P, DIAS-N) Redundancy through Systems Seismic Design Independent Verification Independent Validatio't Independent QA Complete Configuration Control l Corrective Actions Program Multi-task CPUs Minimum Interrupts Deterministic Performance Off-line Changes Only Restricted Data Communications with Isolation ETSafety (Pr5, E5F-CC5)- tdon-1E./cordyc\ (.PCS, Pruess-cC5') Channelized Redundancy do b A

                                                                                   "^d"*"6                            e i %d)

Seismic - No w. Athmti. (wih e(cepbin of MCPA%P moun Independent Verification _ g, cob gec;.htid6 n Independent QA ._ wid VW6s Adn Complete Configuration Control . uk .4a\c. c Pv4 w KinMe TaskfPUc !

                                                         -     nMm m InbrNpk1 l                                                          - l\ lob Cem f 4 he     TJek e NNn'L4b1 Approved Design Afstenini Instrumentation and Contral - Cn GW [kgn CLb                           l' age 7.7 1
                                                           . Naq. y thrickp ' DADA         MWluolja hb yta (PfbuS5 - CCb) l                              - ud a o,a.: w n A r , ... .. . . ;, A , , .m t a ,ta. . t e c si          .

System 80+ Design ControlDocument m No Interrupts ~ Deterministic Design N j Off-line Changes Only ' l Read-on'y Data Communications with Isolation i These systems are provided as an integral part of the Nuplex 80+ Advanced Control Complex (ACC). As such, they provide the operating staff the ability to monitor the plant's operating status, change its operating mode and take those actions necessary to maintain the plant within its design basis for all normal modes of plant operation. The ACC systems and equipment described below include the control systems, main control panels and monitoring systems. Non-lE Category 1 and Category 2 control and monitoring systems which interface with Class IE Safety systems are designed such that creditic failures in the control and monitoring systems will not impact the operation of IE safety systems. Interfaces between these systems employ isolation devices to maintain electrical independence between channels. Isolation devices are qualified for design base events including seismic, environmental, electromagnetic interference (EMI) and electrical fault isolation. Where IE safety related transmitters and signal conditioning devices provide parameters for control and/or monitoring, signal isolation is applied between the safety systems and the control and monitoring systems. Signal validation is performed within the control systems as described in Section 7.7.1.1.13. The DIAS and DPS monitoring systems receive data from both Class IE and non-Class IE systems via qualified fiber-optic network interfaces to maintain isolation. Main Control Panels and the Remote Shutdown Panel are designed to support human-machine interface ! devices for each Class IE safety channel as well as non-Class IE control channels. Low energy circuits are used and all panel mounted control and monitoring devices are isolated such that credible electrical faults originating in one channel will not affect other channels. Control panels are designed to maintain I structural integrity, such that no control room missile hazard will occur as a result of a seismic event. 1 7.7.1.1 Control Systems ! The general description given below permits an understanding of the reactor and important subsystem control methodology. l The design reactivity feedback properties of the NSSS will inherently cause reactor power to match the total NSSS load. The resulting reactor coolant temperature at which this occurs is a controlled parameter and is adjusted by changes in total reactivity as implemented through CEA position changes or through ! boric acid concentration changes in the primary cool:mt. The ability of the NSSS to follow turbine load changes is dependent on the ability of the control systems or operator to adjust reactivity, feedwater flow, bypass steam flow, reactor coolant inventory, and energy content of the pressurizer such that NSSS conditions remain within normal operating limits. Except as limited by Xenon conditions, the major control systems described below provide the capability to automatically follow design load changes. Additionally, these automatic systems provide the capability to accommodate load rejections of any magnitude or the loss of one of two operating feedwater pumps. Approved Design Material Instrumentation and Control Page 7.7-2

System 80+ Design ControlDrcument The depth and quality of training shall be at least equivalent to that provided by the diesel generator manufacturer. Ccc cob The COL applicant shall establish a preventative maintenance program which encompasses investigative 0,. ) testing of components which have a history of repeated repair and replacement of those components with

other products of proven reliability.

8.3.1.1.5 Non-Class IE Alternate AC Source Standby Power Supply <r s't,- - The Alternate AC Source (AAC) is a non-Class 1E combustion turbine power sourc

                                                                                                    ^ *g,

(, ' I with Loss of Offsite Power (LOOP) and Station Blackout (SBO) scenarios. T1 v v - independent and diverse from the Class lE standby emergency diesel generators. (j yn i

                                                                                                                                               . A The combustion turbine-generator and supporting auxiliaries are non-safety related, a                                                          !

3d a packaged unit. mounted in a self contained metal enclosure. The AAC facility is i 2 - plant protected area, outside of the turbine missile impact zone, and is surrounded by a fence and intrusion alarm system. The security fence is locked, and monitored in the ! p g j The combustion turbine fuel tank is surrounded by a dike, sized to retain the entire cc S'q_ ' Additional details of the combustion turbine-generator and fuel storage structures are p 1.2.16.2. The location and arrangement of the AAC components / structures is show h The AAC is sized with sufficient capacity to accommodate either of the following loa O. g 3

Both sets of X and Y Permanent Non-safety loads for a worst case unit V' "

shutdown; or xt e ,one set of Permanent Non-Safety loads and one set of a Safety Division's loads for a worst case

               ' unit shutdown to cold shutdown and/or Design Basis Accident as indicated below:

1. Permanent Non-Safety X with Division I only, or, A

2. j permanent Non-Safety Y with Division II only. I The AAC is not normally nor automatically directly connected to any Class IE Safety Load Division. l However, it can be manually aligned to power one Safety Load Division via one Permanent Non-Safety ! Bus, to accommodate an emergency diesel generator failure or out-of-service condition. The AAC is ' provided with a continuous rating capacity margin of at least 10 percent to compensate for load growth. Those security. loads which require an uninterrupted source will be on a UPS in a secure protected area { of the plant. . 8.3.1.1.5.1 AAC Starting and Loading The AAC is designed to start automatically within two minutes from the onset of a LOOP event. Itis then available for loading if either of the 4,160V Permanent Non-Safety Load Buses X and Y become de-energized. Automatic connection and sequentialloading of the X and/or Y permanent non-safety loads will occur utilizing a sequencer design similar to that described in 7.3.1.1.2.3. Approved Design Material- Doctric Power Page 8.3-13

4 1 Systzm 80 + Drsign controlDocument 4 The first level of undervoltage protection is a set of definite time undervoltage relays with inherent time delays of approximately 10 cycles to ride out power system transients. The initiation of any two of the first level undervoltage relays within a single switchgear will initiate a residual voltage transfer to Preferred Interface 11 by tripping the switchgear main breaker. The transfer is completed when an instantaneous undervoltage relay fed from voltage transformers located on the bus indicates the bus voltage has decayed to 25 %. If voltage has not been restored after a second time delay of approximately 10 seconds, the AAC will start. If voltage has not recovered after the AAC is up to speed and voltage ! (approximately 2 minutes), (1) the main breakers for both permanent non-safety switchgear are tripped. l (2) load shed occurs, (3) AAC breakers close on both switchgear and (4) sequencing begins. Note: One division's Class lE safety buses can be fed from the AAC as described in Section 8.3.1.1.5. This connection is done manually through the permanent non-safety bus. The AAC, permanent non-safety, and Class IE bus circuit breakers are interlocked such that the AAC and Class 1E diesel generator cannot be interconnected. The initiation of any two of the second level instantaneous undervoltage relays within a single permanent non-safety switchgear starts a timer to establish existence of a sustained undervoltage. This timer initiates an alarm in the control room. The operator can re-establish voltage by manually transferring to the alternate offsite source. Failure to restore voltage to an acceptable level after a second time delay starts the AAC and restore power to the permanent non-safety loads. The dropout for the first level of undervoltage protection for the Class IE distribution system is set at a level below minimum voltage during motor starting. Its associated time delay shall be set to ride out power system transients and yet initiate action in a time consistent with the accident analysis (approximately 10 cycles). The dropout for the second level of undervoltage protection for the Class IE distribution system is set at a level above the minimum voltage that shall allow proper operation of Class IE loads with the worst case line-up and minimum switchyard voltage. Its associated first time delay is set to establish existence of a sustained undervoltage (something longer than a motor start - approximately 10 seconds). The second time delay is limited such that the connected Class IE equipment shall not be damaged. The dropout for the first level of undervoltage protection for the permanent non-safety distribution system is set at a level below minimum voltage during motor starting. Its associated time delay is set to ride out , power system transients (approximately 10 cycles). ! l The dropout for the second level of undervoltage protection for the permanent non-safety distribution system is set at a level above the minimum voltage that allows proper operation of permanent non-safety loads with the worst case line-up and minimum switchyard voltage. Its associated time delay is set to establish existence of a sustained undervoltage (something longer than a motor start - approximately 10 seconds). The second time delay is limited such that the connected equipment is not exposed to prolonged voltages below the equ pment rating. - The above settings will be established based on the voltage studies listed below and specific equipment requirements. Voltage studies will be performed by the COL applicant in conformance with Section 3 of Branch h Technical Position PSB-1. The results will then be verified by test as described in Section 4 of Branch Tl Technical Position PSB-1. These voltage studies will be used to determine the relay pickup and time delays of all levels of undervoltage protection described above. Approved Design Material- Electric Power Page 8.3-17

System 80+ Design ControlDocument l l Table 8.3.1-2 Division I Class 1E Loads (Cont'd.) Component Equiv. Loss of Offsite DBA/ LOOP Equipment Component Estimated Motor Load Power Load Load (Note A) No. Per Volts HP pf Efficiency KW Safdy Safdy Safdy Safdy Bus (Note K) Bus A Bus C Bus A Bus C Essential Chilled 1 480 13 BHP 0.9 0.9 10.8 0 0 0 0 Water Pump 1A l (Note O) j RCP Seal Injection 1 480 3 BHP 0.9 0.9 2.5 0 2.5 0 0 l Pump Room i Ventilation Fan Div i Backup 1 480 200KW l.0 - 200 0 200 0 0 Pressurizer Heaters Subtotal leadings for Manual lead Sequence Group: 110 371 130 171 t Total Diesel Load on LOOP Ireluding Manual lead (Note L) 3766 Total Diesel lead on LOOP Excluding Manual Load (Note L) 3208 ! Total Diesel lead on DBA/ LOOP including Manual lead (Note L) 5503 Total Diesel Load on DBA/ LOOP Excluding Manual lead (Note L) 5202 Division I Class IE Loads Diesel Generator lead Sequencer Assumptions and Information: j Note A ((The COL applicant will perform final sizing calculations for the Class IE emergency diesel generators.))l Note B For Station less-of-Offsite-Power or DBA/ LOOP event, lead Sequence Group A is energized immediately upon closure of the Diesel Generator Breaker for Safety Bus A. Note C Although these loads are connected to the Emergency Buses, they are not required once the Diesel Generator is l operating, therefore, they are considered to be zero for purposes of Diesel Generator sizing. Note D lead Sequence Groups I and 2, Safety injection Pumps 1 and 3, will normally be activated by the sequencer to provide Direct Vessel Safety Injection. The Safety Injection System (SIS) requirements specify flow to the vessel within 40 seconds after an Safety injection Actuation Signal (SIAS). This assumes a 20 second Diesel Generator start time, a 2.5 second Safety injection Pump 1 sequence load time (6.5 seconds for Safety injection Pump 3) and a 10 second interval to provide actual SIS flow to the Direct Vessel Injection nozzles. r{NC G to a d 3c ,a ,v w empC coaah k rbV h.th pwnc) { fun; sh (% C. 1 COL information item: see DCD Introduction Section 3.2. Approved Design Material Electric Power Page 8.3-56

         .     ~-- . . . . . .- .                  - . . - -.                        -   .     -                                     .

j SE\T BY:DE&S :11-21-M : 17:07 ; Dl3E ENGR & SRVS4 ;# 8/14 Systeers 80+ 3 Danton coaerd oocament i l Table 8.3.1-3 Division H Clahs IE Imade l I l r' - pn===r Equiv. Lees of Offsite DBA/ LOOP

Equapanent Comiposent Fae-=e e Motor Imewi Power Imed Imad
(Note A) No. Per Voks HP pf EF"hy KW j Bas > K)

Wy Sef e

garg g ,,,

1And ' . + Group B (Note B) 544 S Division 11 Diesel 1 480 5 HP 0.9 0.9 4' 4 0 4 0

Generator Building j Sump Pump 2A 1 Division U Diesel 480 0.9 1 15 KVA - 13.5 13.5 0 13.5 0 i

Genermor 1-480/208/120 Power

Panelboard j Divisson H Diesel 1 480 120 HP 0.9 0.9 90 90 0 90 0
Generator Building
Ventdmion Fan 2A Division 11 Diesel 1 480 0.25 HP 0.9 0.9 0.2 0.2 0 0.2 0
Crankcase Blower

! Division 11 Diesel 480 7.5 HP 0.9 0.9 2 1 5.6 0 0 0 0 Generator Prelube Note C ] Pump j Division H Diesel 1 480 2 HP 0.9 0.9 1.5 0 0 0 0 Genermor Engine Note C i Jacket Waser Keep ' Warm Purap Division 11 DM 1 480 75KW 1.0 - 75 0 0 0 0 s Generator Jacker Note C Water Hemer 1 Division H Diesel 1 480 3 HP 0.9 0.9 2.3 0 0 0 i 0 Generator Bamng Note C ., Gear Motor j Division 11 Diesel 480 ' 7.5 K W 1 1.0 - 7.5 0 0 0 0 j Genermor Lube Oil Note C Hemer d Division 11 Diesel 1 480 4.6KW l.0 4.6 0 0 0 0 l } Generator Space Note C a Heater Channel B 125V 1 480 45.5 KVA O.9 40.9 40.9 0 40.9 0 Vital Rattery { Chargers for IAC j Contml Room and 1 480 12.5 KW I.0 - 12.5 12.5 0 12.5 0 j Remote Shutdown Panel Division U

Lighting j

Asuwowser Desqrn assearw assa avwar mg,, a 3.st 4 4

l l l System 80+ Design ControlDocument Table 8.3.1-3 Division II Class 1E Loads (Cont'd.) Component Equiv. Loss of Offsite DBA/ LOOP Equipment Component Estimated Motor Load Power Load Load (Note A) No. Per Volts HP pf Efficiency KW Safdy Safdy Safdy Safdy Bus (Note K) Bus A Bus C Bus A Bus C Subtotal Loadings for Manual lead Sequence Group: 110 371 130 171 Total Diesel Load on LOOP Including Manual Load (Note L) 3688 Total Diesel Load on LOOP Excluding Manual load (Note L) 3207 Total Diesel lead on DBA/ LOOP Including Manual Load (Note L) 5503 Total Diesel Load on DBA/ LOOP Excluding Manual Load (Note L) 5202 Diesel Generator Load Sequencer Assumptions and Information Note A ((The COL applicant will perform final sizing calculations for the Class IE emergency diesel generators.))3 b{ e -l Note B For Station less-of-Offsite-Power or DBA/ LOOP event, lead Sequence Group B is energized immediately upon closure of the Diesel Generator Breaker for Safety Bus B. Note C Although these loads are connected to the Emergency Buses, they are not required once the Diesel Generator is operating, therefore, they are considered to be zero for purposes of Diesel Generator sizing. Note D lead Sequence Groups I and 2, Safety injection Pumps 2 and 4, will normally be activated by the sequencer to provide Direct Vessel Safety injection. The Safety Injection System (SIS) requirements specify flow to the vessel within 40 seconds after an Safety injection Actuation Signal (SIAS). 'lhis assumes a 20 second Diesel Generator start time, a 2.5 second Safety Injection Pump 2 sequence load time (6.5 seconds for Safety injection Pump 4) and a 10 second interval to provide actual SIS flow to the Direct Vessel Injection nozzles. Note E lead Sequence Group D consists of 480V loads powered on Safety Bus D. 8 COL information item: see DCD Introduction Section 3.2. Approved Design Material- Electric Power Page 8.3-67 I

System 80+ Design ControlDocument Table 8.3.1-4 Electrical Bus Loads (Cont'd.) Component Load (KW)N 13.8 KV Non-Safety Bus Y 1: RCP Switchgear 8217 Condensate Polisher Pump 1824 Circulating Water Pump 2940 Circulating Water Pump 2940 Circulating Water Pump 2940 13.8 KV Non-Safety Bus Y-1 Total = 18,861 KW 13.8 KV Non-Safety Bus Y-2: RCP Switchgear 8217 l FW Booster /FW Pump 16,007 I Condensate Pump 3275 Circulating Water Pump 2940 13.8 KV Non-Safety Bus Y-2 Total = 30,439 KW l l 1 l I l 1 l l i l Ill ((The COL applicant will perform final sizing calculations using actual procured equipment.))' 12) S ngle component capable of being powered by either X or Y Non-Safety Buses. 133 Main Pressurizer Heaters capable of being powered from X or Y Non-Safety Buses. 8 COL information item: see DCD Introduction Section 3.2. Approved Design Material Electric Power Page 8.3-75

SyTtem 80+ oesign controlDocument j e/ Table 8.3.2-3 Class 1E 120V AC VitalI&C Power Systems Loadst (Cont'd.) LG Load on Bus (KVA)#y Description Severe DBA SBO Accident j Channel D DIAS-N 14.0 7.0 7.0 PPS 8.0 8.0 8.0 APC (Safety) 1.0 1.0 1.0 ESF-CCS 5.0 4.5 2.5 Holdup Volume Flood Valve - - 0.5 l Control Room Air intake Radiation Monitor 2.0 - -

                                                                                                                                               ,l Total Load        30.0        20.5            19.0           l w                           -

1 U Co L- i (l1 The COL applicant will perform final sizing calculations using actual procured equipment. 2-l 1 1 1 Abbreviation Description l DIAS-P Discrete Indication & Alarm System - Channel P DIAS-N Discrete Indication & Alarm System - Channel N l PPS Plant Protection System APC Auxiliary Process Cabinet ESF-CCS Engineered Safety Feature-Component Control System l l l l Approved Design Materiel

Electric Power Page 8.3-81

l System 80+ Design Control Document Table 8.3.2-4 Class IE DC Vital Power System Loads (Cont'd.) Load on Bus (amperes)DI Description Sever - i l DBAzi SBO '8 Acciden Division 2 Class IE Switchgear Breaker Tripping 80.4 n2: _ _ Inverter 36.7 - 20.8DS Diesel Gen. Control & Field Flash 69.0 l'1 -- -- Containment Equipment Hatch Trolley -- - 37.20$1 Bus Peak Total: 117.1 58.0 Channel B Class IE Switchgear Breaker Tripping 80.4n21 Inverter 225.0 173.3l71 146.7DS Reactor Trip Switchgear 2 . 5 151 - -- Class IE Switchgear & Load Center 14.7 - - Control Atmospheric Pump Valve 4 7. 8153 4 7. 81'l -- EFW Valve EF-100 6 . 6 161 6.6l'1 - EWF Trip & Throttle Valve #2 2 . 2 151 -- - EFW Steam-Driven Pump #2 Turbine 1.0 1.0 - Governor Control SDS Valve RC-408 19.2n u -- 19.200 Control Room Intake Isolation Damper 3.0 D*l -- 3.0 V'l Bus Peak Total: 330.7 228.7 168.9 Channel D Inverter 250.0 1 7 0. 8171 158.30" Reactor Trip Switchgear 2 . 5 151 - -- Class IE Switchgear & Load Centcr 14.7 -- - Control Atmospheric Dump Valve 47.8tsj 47,gini _ EFW Valve EF-104 9 . 9 161 9.91'l - SDS Valve RC-406 38.400 - 38.4';4 Control Room Intake isolation Pamper 3.00*l - 3.00'l Bus Peak Total: 363.3 228.5 199.7 Notes: VI ((The COL applicant will perform final sizing calculations using actual procured equipment.))' col g-l t23 Imads for Design Basis Accident (DBA) for which the batteries are sized to maintain power supply for at least two hours. COL information item: see DCD Introduction Section 3.2. Approved Design Afsterial- Electric Power Page 8.3-81

                                                                                                     & cxst* 3
                                                                                                    ,.eiis sem. c ** ter- C &
                                                                                                     / n- M ri         i 1/o 5 System 80+                                                                     Design ControlDocument       ~p n credible leakage from higher pressure sources. The discharge of these relief devices is routed to a suitable location so that personnel and other nuclear safety related equipment are adequately protected.

ASME III Class 3 requires a minimum size of 3/4 inch for these thermal relief valves. Pressure relief is provided for each reactor coolant pump to protect against the potential overpressurization of the CCWS due to a reactor coolant pump high pressure seal cooler (oithrottleW g -/~cooiebrube rupture. The pressure relief is sized to accept the maximum expected in-leakage from a reactor coolant pump high pressure seal cooler (or throttle seal cooTe) tube rupture. The pressure relief discharge is directed to the containment floor drain sump /which is within the holdup volume. Electric motor operated valves are located on the component cooling water supply and return lines to each reactor coolant pump. These valves can be used to isolate the in-leakage due to a reactor coolant pump high pressure seal coolerhhrottle seaTc5oTe} tube rupture. in case of a major leak in one of the CCWS divisions, the affected division is removed from service and the other division is utilized. Water quality design parameters applicable to the CCWS are given in Table 9.2.2-1. 9.2.2.2.1 Component Descriptions Table 9.2.2-4 contains the component design parameters for the major components. Each component is described in the following subsections. The active valves are described in Section 9.2.2.2.1.9 and listed in Table 9.2.2-5. 9.2.2.2.1.1 Component Cooling Water IIcat Exchangers The CCW heat exchangers are designed to meet specific site conditions. A horizontal shell and tube heat exchanger is discussed in the following sections, however a plate type heat exchanger may be substituted. Sites selecting the plate type heat exchanger shall provide strainer protection against debris or arrangements which allow backflushing on the service water side. Four component cooling water heat exchangers are provided, two per division, to handle the essential and non-essential cooling requirements. The heat exchangers are sized to provide cooling water at no greater than 105'F during normal operation and at no greater than 120*F during shutdown or post-LOCA operating modes. Each operational mode requires a different alignment of component cooling water heat exchangers. These requirements are listed below: Normal Power Operation - 1 HX per division Normal shutdown (24 hours) - all 4 HXs Safety grade shutdown (24 hours) - 2 HXs required in a single division Post-LOCA - 1 HX in either division During normal power operation as cooling requirements increase, the additional heat exchanger in a division may be needed. Valves CC-106, CC-107, CC-108, CC-109, CC-206, CC-207, CC-208, and l Awmd Design Material- AnHiery Systems p>,y ,y,g.yy

i System 80+ Design ControlDocument l l [ Each division of the CCWS provides cooling for the following redundant safety related components.

Shutdown cooling heat exchangers (2 total,1 per division).

e Shutdown cooling mini-flow heat exchangers (2 total, I per division). e Safety injection pump motor coolers (4 total,2 per division).

Containment spray heat exchangers (2 total, I per division).

e Shutdown cooling pump motor coolers (2 total, I per division). e Containment spray pump motor coolers (2 total, I per division). e Containment spray mini-flow heat exchangers (2 total, I per division). e Component cooling water pump motor coolers (4 total,2 per division).

Spent fuel pool cooling pump motor coolers (2 total, I per division).

e Motor driven emergency feedwater pump motor coolers (2 total, I per division). l i e Diesel generator engine jacket wa.er cooler (2 total, I per division).

Essential chilled water condensers (2 total, I per division).
Spent fuel pool cooling heat exchangers (2 total, I per division).

l The non-essential components are divided between the two divisions of the CCWS. The split was based l upon (1) creating similar flow and heat load requirements between the two divisions and (2) component i locations. These components are listed below:

Reactor coolant pump (RCP) motor air coolers (4 total,2 per division).
RCP motor oil coolers (4 total,2 per division). (Note: Each set of RCP motor oil coolers contains one upper and one lower bearing oil cooler).

i e RCP oil coolers (4 total,2 per division). RCP seal coolers (8 total,4 per division). - e RCP high pressure coolers (4 total,2 per division). ( e Letdown heat exchanger (1 total, serviced by division 2).

Charging pump mini-flow heat exchanger (2 total, I per division).

e Sample heat exchangers (14 total, serviced by division 2 - 8 Primary Sample Heat Exchangers and 6 Steam Generator Primary Sample Heat Exchangers). Approved Design Meterial Auxhliary Systems Page 9.2-26 l

System 80+ Design ControlDocument l l

16. Diesel generator engine starting aftercoolers I A, IB,2A, and 2B outlet temperatures.

17. Component cooling water pump motor coolers I A, IB, 2A and 2B outlet temperatures.

18. Essential chilled water condensers 1 and 2 outlet temperatures.

Control Room Indication l

l Control room indication is provided for component cooling water heat exchangers I A, IB, 2A, and 2B inlet and outlet temperatures.

Test Points '

Temperature test points are provided for the following process temperature parameters:

1. CCW heat exchangers I A, IB, 2A, and 2B inlet and outlet temperatures.

2. Sample heat exchanger outlet header temperatures.

! 3. Reactor coolant pump coolers inlet header temperature.

4. Reactos coolant pumps IA, IB,2A, and 2B high pressure cooler outlet temperatures.

h Reactor coolant pumps I A, IB,2A, and 2B seal cooler outlet header temperaturp 4 g# Reactor coolant pumps I A, IB,2A, and 2B oil cooler outlet temperatures. (,, . Reactor coolant pumps IA, IB, 2A, and 2B motor lower bearing oil cooler outlet temperatures. 7 8. Reactor coolant pumps IA, IB,2A, and 2B motor air cooler outlet temperatures. g' . Reactor coolant pumps IA, IB, 2A, and 2B motor upper bearing oil cooler outlet j temperatures. I o) 0. Charging pump mini-flow heat exchangers 1 and 2 outlet temperatures, to 1. Charging pump motor coolers 1 and 2 outlet temperatures. ! g 2. Instrument air compressor I A, IB,2A, and 2B outlet temperatures. 1

           ;y if.       Normal chilled water condensers IA, IB,2A, and 2B outlet temperatures.

Controls

1. Component Cooling Water Heat Exchanger Outlet Temperature Component cooling water heat exchanger bypass control valves, CC-100, CC-101, CC-200, and CC-201, are modulated to maintain a 95F minimum heat exchanger outlet temperature.

Approved Destgrr Metenini Ausniney Systems Page 9.2-34

__ __ ~~__ __ _ _ _ _ ___ _ _ _ _ . _ ___ System 80+ Design ControlDocument

2. Letdown Heat Exchanger Temperature Control Letdown heat exchanger valve, CC-244, is modulated to control the letdown heat exchanger outlet temperature on the CVCS side.

3. Charging Pump Mini-Flow Heat Exchanger Temperature Control l Charging pump mini-flow heat exchanger control valves, CC-145 and CC-245, are modulated to control outlet temperature of the CVCS side of the heat exchanger.

Alarms Component cooling water heat exchanger high and low outlet temperature is alarmed in the control room.

9.2.2.5.3 Flow

Local Indication Local indication is provided for the following process flow parameters:

                    . Spent fuel pool cooling heat exchangers 1 and 2 outlet flows.

2. Shutdown cooling heat exchangers 1 and 2 outlet flows.

3. Shutdown cooling pump motor coolers 1 and 2 outlet flows.

4. Safety injection pump motor coolers 1, 2, 3, and 4 outlet flows.

5. Containment spray pump motor coolers 1 and 2 outlet flows. j i

6. Letdown heat exchanger outlet flow.

7. Gas stripper outlet flow.

l l 8. Boric acid concentrator outlet flow. l

9. Reactor coolant pumps 1A, IB,2A, and 2B high pressure coolers outlet flows.

l h Reactor coolant pumps 1A,1B,2A, and 2B seal coolers outlet headerjn g E. Reactor coolant pumps 1 A, IB,2A, and 2B oil coolers outlet flows. q , .12. Reactor coolant pumps I A, IB,2A, and 2B motor lower bearing oil coolers outlet flows.

13. Reactor coolant pumps I A, IB,2A, and 2B motor air coolers outlet flows.

     'O            14. Reactor coolant pumps I A, IB,2A, and 2B motor upper bearing oil coolers outlet flows.

g 15. Emergency feedwater pump motor coolers 1 and 2 outlet flows. Asqproved Design historiel- Auxniery Systems Page 9.2 35 I- .. - . .

l Svotem 80+ Design ControlDocument 6 E. Containment spray heat exchangers 1 and 2 outlet flows.

17. Spent fuel pool cooling pump motor coolers 1 and 2 outlet flows. ;

1

18. Containment spray mini-flow heat exchangers 1 and 2 outlet flows.

19. Shutdown cooling mini-flow heat exchangers 1 and 2 outlet flows.

20. Component cooling water radiation monitors 1 and 2 inlet flows, i

21. Sample heat exchangers (each) outlet flows.

22. Diesel generator engine jacket water cooler I and 2 outlet flows.

I

23. Diesel generator engine starting air aftercoolers lA, IB,2A, and 2B outlet flows. j i

24. Component cooling water pump motor coolers l A, IB,2A, and 2B outlet flows.

25. Essential chilled water condensers 1 and 2 outlet flows.

26. Charging pump mini-flow heat exchangers 1 and 2 outlet flow.

27. Charging pump motor coolers I and 2 outlet flows.

28. Instrument air compressor I A, IB,2A, and 2B outlet flows.

29. Normal chilled water condensers I A, IB,2A and 2B outlet flows.

11 30". Makeup water to surge tanks 1 and 2 inlet flows. e Control Room Indication l Control room indication is provided for component cooling heat exchangers I A,1B, 2A, and 2B outlet flows and component cooling water pumps IA,19,2A, and 2B discharge flows.

Test Points Flow test points are provided for the component cooling water heat exchangers I A, IB, 2A, and 2B outlet flows.
Controls The following essential heat exchangers have control valves that modulate their outlet flow.

1. Spent fuel pool cooling heat exchangers 1 and 2: CC-112 and CC-212.

2. Shutdown cooling heat exchangers I and 2: CC-Il0 and CC-210.

Approved Design Materiel Ausniery Systems Page k.2-36

 . .___. . - ~~                      .-             .-       .   . . -    -.       . .    . . - -        .    . -       - - . - - - . .

k Syatem 80+ Destan controlDocument

e. Alarms The following low and high flows alarm to the control room:

1. Spent fuel pool heat exchangers 1 and 2 low and high outlet flows.

2. Shutdown cooling heat exchangers 1 and 2 low and high outlet flows.

3. Shutdown cooling pump motor coolers 1 and 2 low outlet flows.

4. Safety injection pump motor coolers 1, 2, 3, and 4 low outlet flows.

5. Containment spray pump motor coolers I and 2 low outlet flows.

6. Reactor coolant pumps IA, IB,2A, and 2B high pressure coolers low outlet flows.

                           . Reactor coolant pumps I A, IB, 2A, and 2B seal coolers low outlet heasleN_ =[>%.

r[ t Reactor coolant pumps lA, IB,2A, and 2B oil coolers low outlet flows.

4 . J). Reactor coolant pumps I A, IB, 2A, and 2B motor lower bearing oil coolers low outlet flows. a) 40. Reactor coolant pumps I A, IB, 2A, and 2B motor ai(coolers low outlet flows. g A f. Reactor coolant pumps lA, IB,2A, and 2B motor upper bearing oil coolers low outlet flows.

12. Emergency feedwater pump motor coolers 1 and 2 low outlet flows.

13. Containment spray heat exchangers 1 and 2 low outlet flows.

14. Spent fuel pool cooling pump motor coolers I and 2 low outlet flows.

15. Diesel generator engine jacket water cooler 1 and 2 low outlet flows.

16. Diesel generator engine starting air aftercoolers I A, IB,2A, and 2B low outlet flows.

17. Component cooling water pump motor coolers lA, IB,2A, and 2B low outlet flows.

18. Essential chilled water condensers 1 and 2 low outlet flows.

19. Charging pump motor coolers 1 and 2 low outlet flows.

20. Instrument air compressor I A, IB,2A and 2B low outlet flows.

21. Normal chilled water condensers IA, IB,2A, and 2B low outlet flows.

4 j -22. Component cooling water heat exchangers IA, IB, 2A, and 2B low and high outlet flows. Anwoved Desigre nieteniel- Am6ery Systems Page 9.2-37

System 80+ Design ControlDocument

                  @ 23:          Component cooling water pumps I A, IB,2A, and 2B low and high outlet flows.

I

d. K Component cooling water radiation momtors 1 and 2 low outlet flows. !

9.2.2.5.4- Level j 1 e Component Cooling Water Surge Tank Level l Level indication is provided in the control room for component cooling water surge tanks 1 and

2. High level, demineralized water automatic supply, low level, and low-low level alarms are provided in the control room.

A low-low level signal isolates the non<ssential headers and the RCP headers from the remaining portion of the system. j 1 1 e Component Cooling Water Sump Level The component cooling water sumps 1 and 2 water levels are indicated and a high level alarm is provided in the control room. Each coniponent cooling water sump pump.is automatically started at a specified sump level, and the pumps are automatically stopped at sump low level. e Component Cooling Water Heat Exchanger Structure Sump Levels l Component cooling water heat exchanger maintenance and floor drain sumps 1 and 2 water levels are indicated and a high level alarm is provided in the control room. The component cooling water heat exchanger structure sump pumps are automatically stopped a* a low sump level, e Component Cooling Chemical Addition Tanks Level Local level indications are provided for the component cooling chemical addition tanks. l ! 9.2.2.5.5 Radiation Monitors l Radiation monitors are provided downstream of the component cooling water pumps. An alarm is j sounded in the control room if radiation is detected at a preset level above background by one of the i monitors. Component cooling water radiation activity is indicated in the control room. 9.2.2.5.6 Current Component cooling water pump motor current is indicated in the control room. 9.2.2.5.7 Interlocks The component cooling water sump pumps are automatically started when the sump level rises to a predetermined height. At this level in sump 1, valve CC-153 opens; and at this level in sump 2, valve CC-253 opens. The sump pumps pump component cooling sump water to their respective surge tank and are automatically stopped at a preset surge tank level or a sump low level. Valves CC-153 and CC-253 close when their respective sump pumps are automatically stopped. Apnproved Deskrs ateteria!- Ausinney Systems Pege 9.248

I System 80+ Design ControlDocument 1 Table 9.2.2-3 Typical Component Cooling Water System Heat Loads and I Flow Requirements Normal Operation Total Number Number With Heat Load Receiving Total Component Heat Load (E+ 06 How How Div.I Div. 2 mu/hr) Div.I Div. 2 %Pm) f l Shutdown Cooling Heat Exchangers 0 0 0 0 0 0 l Shutdown Cooling Pump Motor 0 0 0 1 1 60 i Coolers Shutdown Cooling Mini-Flow Heat 0 0 0 1 1 320 t Exchangers Safety injection Pump Motor Coolers 0 0 0 2 2 IL Containment Spray Heat Exchangers 0 0 0 0 0 0 ; Contamment Spray Pump Motor 0 0 0 1 1 60 Coolers Containment Spray Min! Flow Heat 0 0 0 1 1 320 Exchangers Component Cooling Water Pump 1 1 0.41 2 2 354 , Motor Coolers Spent Fuel Pool Cooling Pump Motor 1 0 0.62 1 1 80 t Coolers Emergency Feedwater Pump Motor 0 0 0 1 1 60 Coolers : Spent Fuel Pool Cooling Heat 1 0 9.6 1 0 5000 Exchangers Diesel Generator Engine Jacket Water 0 0 0 1 1 2000 Coolers , Diesel Generator Engine Starting Air 0 0 0 2 2 100 l Aftercoolers Essential Chilled Water Condensers 0 0 0 1 1 1620 RCP Motor Air Coolers 2 2 6.44 2 2 1200 RCP Mo'or Oil Coolersu l 2 2 0.612 2 2 192 RCP Oil Coolers ,fc/ 2 2 1.8 2 2 281.6 l M &KA- A. M~& RCP High Pressure Coolers 2 2 0.748 2 2 300 letdown Heat Exchanger 0 1 22.7 0 1 1500 Charging Pump Motor Coolers a 0 0.577 1 1 140 Approved Design Meteriet Aunniary Systems Page 9.2-67

A System 80+ nesian controlDocument Table 9.2.2-3 Typical Component Cooling Water System Heat Loads and Flow Requirements (Cont'd.) Nonnal Operation (Cont'd.) Total Number Total Number With Heat Load Receiving Nw Component Heat Load (E+ 06 Nw (spm) Div.1 Div. 2 mu/hr) Div.1 Div. 2 Charging Purg Mini-Flow Heat 1 0 1.98 1 1 800 Exchangers Primary Sample Heat Exchangers 0 8 4.08 0 8 240 Steam Generator Primary Sample Heat 0 6 -3.24 0 6 120 Exchangers Gas Stripper 0 1 17.6- 0 1 700 Boric Acid Concentrator 0 1 14 0 1 700 Normal Chilled Water Condensers 1 1 24 2 2 12000 instrument Air Compressor Oil 1 1 0.585 2 2 200 Coolers, Intercoolers, Jacket Coolers, and Aftercoolers f 0*7N Total Heat lead per Division 1 = 30fL3i@ E+06 Bru/hr + '50.e') YF Total Heat Imad per Division 2 = 78.9 +06 Btu /hr e T8.9#~1G Total Flow per Division 1 = y gpm t? 5 Total Flow per Division 2 = [ 94y) gpm esm.e mes.s Approved Design Meterie!

Aushiiery Syswns Page 9.2-68

Sy~t:m 80 + Design ControlDocument Table 9.2.2-3 Typical Component Cooling Water System Heat Loads and Flow Requirements (Cont'd.) Shutdown Cooling (Initial) Total Number Total I i Number With Heat Load Receiving How ! l Component IIeat Load (E + 06 , How (gpm) l Div.I Div. 2 Btu /hr) Div.1 Div. 2 Shutdown Cooling Heat Exchangers 1 I 262.2 1 1 26000 l Shutdown Cooling Pump Motor Coolers 1 1 0.222 1 1 60 l Shutdown Cooling Mini-Flow Heat 1 1 1.36 1 1 320 .I Exchangers Safety injection Pump Motor Coolers 0 0 0 2 2 160 Containment Spray Heat Exchangers 0 0 0 0 0 0 Containment Spray Pump Motor Coolers 0 0 0 1 1 60 Containment Spray Mini-Flow Heat 0 0 0 1 1 320 Exchangers Component Cooling Water Pump Motor 2 2 0.82 2 2 354 Coolers Spent Fuel Pool Cooling Pump Motor Coolers 0 0 0 1 1 80 Emergency Feedwater Pump Motor Coolers 0 0 0 1 1 60 Spent Fuel Pool Cooling Heat Exchangers 0 0 0 0 0 0 Diesel Generator Engine Jacket Water Coolers 0 0 0 1 1 2000 Diesel Generator Engine Starung Air 0 0 0 2 2 100 Aftercoolers Essential Chilled Water Condensers 0 0 0 1 1 1620 j l RCP Motor Air Coolers 1 1 3.22 2 2 1200 RCP Motor Oil Coolerstu 1 1 0.306 2 2 192 RCP Oil Coolers ,A 1 1 0.9 2 2 281.6 itCf%eebCeedwA 1 1 0.004 2 2 70.4 E q l RCP High Pressure Coolers 1 1 0.374 2 2 300 Letdown Heat Exchanger 0 1 12 0 1 990 Charging Pump Motor Coolers 1 0 0.577 1 1 140 Charging Pump Mini-Flow Heat Exchangers 1 0 1.98 1 1 800 Primary Sample Heat Exchangers 0 8 4.08 0 8 240 Steam Generator Primary Sample Heat 0 6 3.24 0 6 120 Exchangers Gas Stripper 0 1 17.6 0 1 700 l Boric Acid Concentrator 0 1 14 0 1 700 l Normal Chilled Water Condensers 1 1 24 2 2 12000 Instrument Air Compressor Oil Coolers, 1 1 0.585 2 2 200 Intercoolers, Jacket Coolers, and Aftercoolers Total Heat Load per Division 1 = 149.5826. E+06 Btu /hr / % 66 c'E 7,tal lleat Load per Division 2 = 197.94M E+06 Btu /hr / 97. C9 i M h ' "Sw per Division 1 = 23159 gpm t.S p s. S Tot Ilow per Division 2 = 25909 gpm 266/6 6 Approved Desigru Material- Auxiliary Systems Page 9.2-69 l

System 80+ Design ControlDocument Table 9.2.2-3 Typical Component Cooling Water System Heat Loads and Flow Requirements (Cont'd.) Shutdown Cooling (Final) Total Number Total Number With Ileat Load Receiving How Component IIeat Load (E+ 06 Mow (gpm) Div.I Div. 2 Div.1 Div. 2

 -RC+3eal-Chia.                             O           G             0               ?             ?                         '0. '  %

RCP High Pressure Coolers 0 0 0 2 2 300 Letdown Heat Exchanger 0 1 0.51 0 1 35 Charging Pump Motor Coolers 1 0 0.577 1 1 140 Charging Pump Mini-Flow Heat 1 0 1.98 1 1 800 Exchangers Primary Sample Heat Exchangers 0 8 4.08 0 8 240 Steam Generator Primary Sample 0 6 3.24 0 6 120 Heat Exchangers Gas Stripper 0 0 0 0 1 700 Boric Acid Concentrator 0 1 14 0 1 700 Normal Chilled Water Condensers 1 I 24 2 2 12000 Instrument Air Compressor Oil 1 1 0.585 2 2 200 Coolers, Intercoolers, Jacket Coolers. and Aftercoolers Total Heat lead per Division 1 = 56.2705 E+06 Bru/hr Total Heat lead per Division 2 = 65.3235 E+06 Btu /hr Total Flow per Division 1 = 28464 gpm 181 t'5 6 Total Flow per Division 2 = % pm 24 c)IB 6 Approved Design Material- AuxHiery Systems Page 9.2 71

I System 80+ Design ControlDocument Table 9.2.2-3 Typical Component Cooling Water System Heat Loads and Flow Requirements (Cont'd.) Refueling Operations , Number With Total Number Receiving Total Ileat Load IIcat Load __ Flow How Component niv,1 giv, 2 (E+ 06 Div.1 Div. 2 (gpm) j Btu /hr) Shutdown Cooling Heat Exchangers 1 1 60 1 1 26000 l Shutdown Cooling Pump Motor Coolers 1 1 0.222 1 1 60 Shutdown Cooling Mini-Flow Heat Exchangers 1 I 1.36 1 1 320 Safety injection Pump Motor Coolers 0 0 0 2 2 160 Contamment Spray Heat Exchangers 0 0 0 0 0 0 Containment Spray Pump Motor Coolers 0 0 0 1 1 60 Containment Spray Mini-Flow Heat 0 0 0 1 1 320 Exchangers Component Cooling Water Pump Motor 2 2 0.82 2 2 354 Coolers Spent Fuel Pool Cooling Pump Motor Coolers 1 1 1.24 1 1 80 Emergency Feedwater Pump Motor Coolers 0 0 0 1 1 60 Spent Fuel Pool Cooling Heat Exchangers 1 I 19.lWp 1 1 10000 Diesel Generator Engine Jacket Water Coolers 0 0 0 (21 1 1 2000 Diesel Generator Engine Starting Air 0 0 0 2 2 100 Aftercoolers Essential Chilled Water Condensers 0 0 0 1 1 1620 RCP Motor Air CoolerN' ' O O O 2 2 1200 RCP Motor Oil Coolers01 0 0 0 2 2 192 RCP Oil Coolers 0 0 0 2 2 281.6 .prp utr w f) n n n 7 n4 --- RCP High Pressure Coolers 0 0 0 2 2 300 l Letdown Heat Exchanger 0 0 0 0 0 0 Charging Pump Motor Coolers 0 0 0 1 1 140 l Charging Pump Mini-Flow Heat Exchangers 0 0 0 1 1 800 Primary Sample Heat Exchangers 0 0 0 0 8 240 Steam Generator Primary Sample Heat 0 0 0 0 6 120 Exchangers Gas Stripper 0 1 17.6 0 1 700 Boric Acid Concentrator 0 1 14 0 1 700 Normal Chilled Water Condensers 1 1 24 2 2 12000 Instrument Air Compressor Oil Coolers, 1 1 0.585 2 2 200 Intercoolers, Jacket Coolers, and Aftercoolers Total Heat lead per Division 1 = 53.7085 E+06 Bru/hr Total Heat lead per Division 2 = 85.3085 E+06 Bru/hr Total Flow per Division 1 = 28!59 gp;r. cb 6 823. t3 p Total Flow per Division 2 = 29919 gpm 2c) St3. S p Approved Design Material- Auxiliary Systems Page 9.2-72

System 80+ Design ControlDocument Table 9.2.2-3 Typical Component Cooling Water System Heat Loads and Flow Requirements (Cont'd.) I Design Basis Accident Number With :lotal Number Receiving Total Heat Load 1. cat Load How How Div.I Div. 2 Div.I Div. 2 Btu /hr) Shutdown Cooling Heat Exchangers 0 0 0 0 0 0 , Shutdown Cooling Pump Motor 0 0 0 1 1 60 Coolers Shutdown Cooling Mini-Flow Heat 0 0 0 1 1 320 Exchangers Safety injection Pump Motor 2 2 0.044 2 2 160 Coolers ,.I.M Contamment Spray Heat Exchangers 1 1 10h 1 1 16000 Contamment Spray Pump Motor 1 1 0.222 1 1 60 Coolers Containment Spray Mini-Flow Heat 1 1 1.36 1 1 320 Exchangers Component Cooling Water Pump 1 1 0.41 2 2 354 Motor Coolers Spent Fuel Pool Cooling Pump 0 0 0 1 1 80 Motor Coolers Emergency Feedwater Pump Motor 1 1 0.136 1 1 60 Coolers Spent Fuel Pool Cooling Heat 0 0 0 0 0 0 Exchangers Diesel Generator Engine Jacket 1 1 38.1 1 1 2000 Water Coolers Diesel Generator Engine Starting 2 2 0.0344 2 2 100 Air Aftercoolers Essential Chilled Water Condensers 1 1 6.48 1 1 1620 RCP Motor Air Coolers [s3 0 0 0 2 2 1200 RCP Motor Oil Cooler [ 0 0 0 2 2 192 RCP Oil Coolers 0 0 0 2 2 281.6

        -1tCP                                               'WM7WWN RCP High Pressure Coolers                 0            0             0            2        2           300 Letdown Heat Exchanger                    0            0             0            0        0               0 Approved Design Meterial AuxHiery Systems                                                         Page 9.2-73

l l l Design ControlDocument t System 80+ l Table 9.2.2-3 Typical Component Cooling Water System Heat Loads and Flow Requirements (Cont'd.) Design Basis Accident Number With Total Number Receiving Total Heat Load Heat Load How Flow anPonent Div.I Div. 2 Div.1 Div. 2 Btu /hr) Charging Pump Motor Coolers 1 0 0.577 1 1 140 Charging Pump Mini-Flow Heat 1 0 1.98 1 1 800 Exchangers Primary Sample Heat Exchangers 0 0 0 0 0 0 Steam Generator Primary Sample 0 0 0 0 0 0 Heat Exchangers Gas Stripper 0 0 0 0 0 0 Boric Acid Concentrator 0 0 0 0 0 0 Normal Chilled Water Condensers 0 0 0 0 0 0 Instrument Air Compressor Oil 1 1 0.585 2 2 200 Coolers. Intercoolers, Jacket Coolers, and Aftercoolers Total Heat Load per Division 1 = 134.2427 E+06 Bru/hr Total lieat Load per Division 2 = 131.6857 E+06 Blu/hr l Total Flow per Division 1 = y) gpm it 17 3'6 ! Total 'ilow per Division 2 = 121,39 171 gpm it "4 8, .f.a Notes: i [1] Each set contains one upper and one lower bearing oil cooler. Data applies to the two cooler combination. F2t- F= - P::: :seka pu pu..+ G-u a m divisivu). i. rata iin oyy b c; ;o i: :::: r"~ r~.f...Anm j g. [3] The listed heat load for the spent fuel pool cooling heat exchangers does not give consideration e a single active failure. Under this condition, the heat load on a single spent fuel pool cooling heat exchanger would i be 19.19 x 106 Btu /hr. Likewise, a single active failure coincident with a full core offload would result

in a heat load of 67.25 x 10 6Btu /hr on a single spent fuel pool cooling heat exchanger, i

3 p] This heat load must be carried by each division. ( l l l l Approved Design Material- Auxiliary Systems Page 9.2-74 i i

CESSAR 8%"ic41 cu l TABLE 9.2.3-2 PROCESS MONITORING PARAMETERS B i Demin.f4 Cation Decarbon. Anion Mixed Bed DegasiW Parameter inlet Efflu. Efflu. Efflu. __E f fl u . Efflu. l Sodium C C W Chloride C C l Sulfate W W l Silica C C pH C C Conductivity C C W Pressure C C l Flow C- -- - - -~

                                                                      ~ -
                                                                                .  -g l   .Dissued&gn.. -              -        -    -
                                                                                --.C        l NOTES:                                                                                  ,

1. All parameters listed above are either continuously monitcred (C) or grab sampled on a weekly basis (W).

2. Recorders are provided for all continuous parameters except pressure. A flow totalizer is also provided.

3. Alarms are provided for mixed bed effluent chemistry parameters, low demineralizer flow and degasifier effluent dissolved oxygen.

4. H decarbonators are utilized.

Amendment B March 31, 1988

CESSAR Uha _

2. Non-current site related events, including transportation accidents, oil spills and fires.

3. Credible single failures of man-made structures.

4. Sabotage C1 E. In order t ure the normal cooldown (two division requirements given in Section cooldown) A erformance 5.4.7.1.2.F.A the Ultimate Heat Sink temperature at the SSWS inlet mu'st' remain at or below 95'F throughout the cooldown.

One percent exceedance meteorologic conditions can be utilized in showing acceptability, since this is a performance requirement and not a safety requirement. F. For sites with severe winters, where ice formation of the Ultimate Heat Sink could occur, an analysis shall be

provided showing the function of the Ultimate Heat Sink is l not impaired during winter months. Where required, the intake structures shall be provided with a means of deicing, such as warm water recirculation, to prevent flow blockage of the SSW pump inlets.

G. The site water chemistry for the Ultimate Heat Sink shall be analyzed to determine if a water treatment system is required to minimize corrosion and fouling of the SSWS. l H. Water boundaries that form part of the protected area boundary shall be avoided, if at all possible. 9.2.5.2 System Description The Ultimate Heat Sink described here consists of a single passive independent cooling water pond connected to the SSWS through intake and discharge paths. However, it is recognized that site-specific conditions may require the use of two ponds to meet Regulatory Guide 1.27. The design brackets alternative Ultimate Heat Sinks which may be specified for a particular site if environmental restrictions limit the use of a cooling pond or if an alternative water supply is more reliable. Acceptable alternate ultimate heat sinks are an ocean, a large lake, a large river, a lake and a cooling pond, a river and a cooling pond, or a cooling tower and cooling pond. The cooling water pond is provided with makeup water pumps to maintain level. Water chemistry is maintained by a site-specific water treatment system (i.e., chemical injection). Salinity buildup in a pond is limited by blowdown. The Ultimate Heat Sink will operate for the required nominal 30 days following a postulated LOCA without requiring any makeup water to the source, and without requiring any blowdown from the pond for salinity control. Amendment U 9.2-64 December 31, 1993

System 80+ Deslan CukolDocummt l l 1 I I i l I

l l l C Heat Balance Diagram 3' Mgure 10.1 1 i NIh4 Db'N bP S U k n \* hte hk o"w; 4 e ) c h 7 4 Q ,(- Q',n3,4, wm o e u w.na m hoe 10.14

  .__...m         ._                   . _ _ _ _ _       _               _        . . _ _ _        _           _ ._     __ _ ____

System 80+ Deenpr ControlDocument i

10.4 Other Features of Steam and Power Conversion System i.

l 10.4.1 Main Condenser i j 10.4.1.1 Design Bases i l e The main condenser is designed to condense the low pressure turbine exhaust steam so it can be ! ' efficiently pumped through the steam cycle. The main condenser also serves as a collection point ] for the following: l

1. Feedwater heater drains and vents

) 2. Condensate and Feedwater System makeup j 3. Miscellaneoas equipment drains and vents e The main condenser is also designed to condense up to 55% of the full load main steam flow l

bypassed directly to the condenser by the Turbine Bypass System. The steam is bypassed to the I main conderner in case of a sudden load rejection by the turbine generator or a turbine trip, and j j at plant startup and shutdown as described in Section 10.4.4. The main condenser hotwells serve i

as a storage reservoir for the Condensate and Feedwater Systems with sufficient volume to supply j maximum condensate flow for 5 mmutes. The main condenser is also designed to provide i removal of noncondensible gases from the condensing steam by the Main Condenser Evacuation j System, described in Section 10.4.2. Heat is removed from the main condenser by the ! Condenser Circulating Water System. i a ) 10.4.1.2 System Description } 'Ihe following functional requirements are to be met to ensure a reliable system: /

                                                                                                       -                                      /

e ' l The condenser is designed in accordance with Heat Exchange Institute Standards. The reference - ! plant condenser incorporates a three zone multipressure condenser. - of two parallel tube bundles to permit maintenance and cleaning during operation. The ' circulating water system is routed to each of three condenser shells in a series configuration. j Tubing is of commercially available lengths. The design does not preclude shop pre-fabrication. l , i Representative main condenser operating parameters for a natural draft cooling tower site are j provided in Table 10.4.1-1. The actual design and operating parameters will be determined on / ] a site specific basis. A schematic drawing of a typical main condenser is provided in Figure j s i, 10.4.1-1. ] 1 N

The condenser tube material is as specified in Section 10.3.6.2. !

4 j e Tube gauge with' stainless steel is not thinner than 22 BWG. Tube gauge with titanium is not thinner than 23 BWG. Condenser design precludes or minimizes steam impingement forces on the candenser tubes for normal operation and turbine bypass valve quick opening events. Tube { support plates are designed to minimize tube vibrations. j [ (owh\ hg MmM^ , 4ee. bcb MLchrn DecAirn '$.4. z - ='onon unauw rower comessen n ye to.s.

                                                                                            .           =           ..

              . _ _ -         -..         -           -                        -         -       _ _ -          ~ .

System 80+ Deslan ConkelDoewnent e Provisions for chemical injection into the condenser for biofouling control is included in l accordance with site-specific requirements and applicable regulations. e Means are provided to protect the tubes from pitting during periods of condenser shutdown. e Tube sheets are specified as stated in Section 10.3.6.2. e Double tube sheets or welded tube to tube sheet joints are provided. e Leak detection trays are included at all tube to tube sheet interfaces. Provisions for early leak detection are provided at tube sheet trays and in each hotwell section. To allow for leak detection and location, the hotwell for each condenser shell is equipped with conductivity cells located under each tube bundle. e The condenser is designed to deserate the condensate during startup and normal operation. The design also deserates any drains which enter the condenser. e 'The condenser and circulating water system are designed to permit isolation of a portion of the tubes (segmented condenser) to permit repair of leaks and cleamng of water boxes while operating at reduced power. e The condenser is capable of being filled with water for a hydrotest. Provisions are made to allow drainmg and cleamng of the hotwell. e A stainless steel expansionjoint and a water seal trough between the condenser and the turbine are provided. 1 e [An automatic condenser cleaning system is provided.)) ~ e Heater shells and piping installed in the condenser neck are located outside of the turbine exhaust steam high velocity regions and within the limits specified by the turbine supplier. Internal piping is as short and straight as possible and all steam extraction piping slopes downward toward the heater shells.

Sections of heater shells and piping that are located inside the condenser and normally operated with a fullload inside temperature of about 194*F or more shall be lagged. The lagging is made of stainless steel at least 1/16-inch thick and is designed consistent with proven practice.

e The condenser neck fluid design is based on air tests, modelling the steam flow path from the low-pressure turbine exhaust hoods to the condenser tube bundles. The test model accounts for the condenser neck heaters and associated piping and for the neck major structural elements, lines and baffles. The tests cover all major operating modes including operation with steam bypass dump and operation with one tube bundle out of service. e The change in liquid inventory in the steam generators, as plant load changes, is considered in designing the Condensate System and sizing the condenser hotwell. On a steady state basis, the steam generator mass decreases by approximately 103,500 pounds per steam generator (207,000 pounds total) between 0 percent and 100 percent load. L p.a.

i Ccaaplua. ..)

e r % ar.enw. re.w comw.A. p.e. ro.4a

Svatem 80+ Deslan ControlDocument Refer to Section 7.7.1.1.12 for a description of the Process-Component Control System which provides ! applicable non-safety remote monitoring and controls from the main control room. 10.4.6 Condana* Cleanup System 10.4.6.1 Design Basis The Condensate Cleanup System (CCS) is an integral part of the Condensate System. The CCS is designed to remove dissolved and suspended impurities which can cause corrosion damage to secondary system equipment. The CCS also removes radioisotopes which might enter the system in the event of a primary to secondary steam generator tube leak. The condensate polishers will also be used to remove impurities which could enter the system due to a condenser circulating water tube leak. 10.4.6.2 System Description The Condensate Cleanup System utilizes side stream, full flow condensate polishers located downstream of the condensate pumps. Iead cation beds with downstream deep bed, mixed resin, ion exchangers are utilized to obtain the advantage of their larger capacity in the event of the inleakage of impurities, to reduce the probability of resin discharge to the feed system due to failure of resin retention elements, and to simplify system operation. [ Representative condensate cleanup system design and operating parameters are listed in Table 10.4.6-1. The Condensate Cleanup System is shown in Figure 10.4.6-1.]f The following functional requirements are to be met to ensure a reliable system. e The Condensate Cleanup System is sized to meet the chemistry requirements for continuous operation specified in Section 10.3.5 while operating with a condenser leak of 0.001 gpm and to i maintain water quality during an orderly unit shutdown (not longer than 8 hours) with a leak of 0.1 gpm regardless of the type of cooling water. e The number and sizing of the polisher vessels are such that the functional requirements can be met while permitting the replacement or regeneration of resin in one ion exchanger at a time.

Plant features are provided to facilitate replacement of ion exchange resin.

e Resin traps are installed downstream of each ion exchanger to remove resin fines. l e Flow rates through the polisher vessels during normal full power operation are 40 gpm/ft or less. 2 In addition, a minimum flow rate is specified by the manufacturer to prevent channeling. ' Minimum bed height is 3 ft. e The polisher vessels' outlet lines are fitted with individual flow regulating valves. e The system design permits full flow recirculation and return to condenser hotwell deaerating sections through each polisher vessel for cleanup and verification of resin bed performance after resin replacement and prior to alignment within the system.

Polisher vessel isolation and recirculation valves are designed to permit slow, controlled opening to minimize hydraulic surges on the resin bed.

y 7,y f *v trnn & Agurowed Deep Afssonist Poever Conversion Aspe 70.4 f6

System 80+ Dennen contrar Document All EFW System components are located in Seismic Category I structures which also protect the components from external environmental hazards. All piping and components essential to EFW operation are designed to Seismic Category I standards as described in Section 3.7, and are designed to accommodate, located to protect against, or protected from internal flooding and internal missiles as discussed in Sections 3.4 and 3.5. Steam binding of the EFW pumps (GSI-93) is mmimized by the following design and operational features: e The temperature sensor located between the EFW flow control valve and the isolation valve on each subtrain is continuously monitored and audibly alarmed in the control room. Provisions are made such that in the event of loss of control room indication, the sensor can and will be monitored locally. Operating procedures, which are the responsibility of the COL i Applicant, will present the requirements for local monitoring. At a muumum, readings shall be recorded at least once a shift, and before and after each EFW pump run. e Ihe EFW system is designed to avoid steam binding of the EFW pumps by continuous system _ venting through the EFW storage tanks and by the use of normally closed isolation valves upstream of the interface with the Main Feedwater system: however, in the event that steam binding of the EFW pumps does occur, the control room alarm associated with the temperature hi sensor dirn=I above will signal the plant operator to vent the EFW procedures developed by the COL Applicant will prescribe this action.;l w t h I g, 10.4.9.4 Inspection and Testing Rw ' ' During fabrication of the EFW components, tests and inspections are performed and documented in #l accordance with code requirements to assure high quality construction. As necessary, performance tests of components are performed in the vendor's facility. The EFW System is designed and installed to permit in-service inspections and tests in accordance with ASME Code Section XI. 10.4.9.4.1 EFW System Perfon= mace Tests Prior to initial plant startup, a comprehensive performance test, as detailed in Section 14.2, will be performed to verify that the design performance of the system and individual components is attained. 10.4.9.4.2 Reliability Tests and Inspections e System Level Tests , After the plant is brought into operation, periodic tests and inspections of the EFW components and subsystems are performed to ensure proper operation. The scheduled tests and inspections are w=ry to verify system operability, since during normal plant operation, EFW components are aligned for emergency operation and serve no other function. The tests defined permit a complete checkout at the component level during normal plant operation. Satisfactory operability of the complete system can be verified during normal scheduled refueling shutdown. The complete schedule of tests and inspections of the EFW System is detailed in Chapter 16. c:.0,

l ('COUnhndo%* ')

J Anrod Des &n aissada!- Power Conversion Page 10.442

1 Syotem 80+ Design ControlDocument l l Table 10.4.1-1 Representative Main Condenser Operating Parameterb y 4 L( l ((The following condenser operating parameters (re representative of typical timated condenser operating conditions based on circulating water system Ionditions as noted below an a single natural draft cooling tower utilized as the circulating water syst ling source. The actual design and operating parameters will be determined on a site specific basis.

                                                                               >           0L Description                                                  ParametedM l

_- u . . .. . p Inlet Circulating Water Temperature (*F) 04o*In d /fnadu M #f- 70 / c)5 - 05 /. f Condenser Data for Total of nree Shells '! Circulating Water Flow (gpm) 600,000 l 1 JQ Approximate Terupu.ims Rise (*F) 29 Tube Material / Gage Titaniumb3 l l Maximum Tube Velocity (ft/sec) 9 l Tube Diameter (inches) 7/8 Number of Passes (3 shells in series) 1 Shell Backpressure (in. Hg) (Momina\ / Madmum) k( .A, [) N Shell 1 I I.k 14/77 [2.7 S uw- , Shell 2 2.3k 2.'s/ 3 5 f 3.5 l Shell 3 1 2.95) 7.95/4.5\4.5 Approximate Tube Length (ft) F[) She!!! 35 q , Shell 2 45 Shell 3 55 Heat Rejection Rate (million BTU /hr) 8600Nh A h 9 i kN Natural Draft Cooling Tower Site i (y) ~p f.N.* > QConcephl "De4Qn - } C9/ Condenser tube material will be selected based on site specific conditions to meet the requirements specified in Section 10.3.6.2. Approved Design Meteriel . Power Conversion Page 10.4-46

Syntem 80+ Deskn Control Document Table 10.4.5-1 Representative Circulating Water System Operating Paramete f ((The following circulating water system operating parameters are representative of tvoical i ted ] operating conditions. These parameters are based on the circulating water system conditions and j bounding meteorological conditions as noted below and utilizing a single natural draft cooling tower as i the circulating water system cool' source. The actual design and operating parameters will be j determined on a site specific basi . [ 1 Description ParameterN Total Cin:ulating Water System Flow (gpm) 600,000 l Flow / Pump (gpm) 150,000 1

;      Circulating Water Condenser inlet Te+4uie ('F)                                 95 h
!      Circulating Water Condenser Outlet Temperature (*F)                            124 Cooling Tower Range (*F)                                                       29 j       Cooling Tower Approach to the Wet Bulb (*F)                                     15                                               l Wet Bulb (*F)                                                                   81 ,   [                                         j

$ Relative Humidity (%) 70 ][ ] l

                                                                                             \                                          l l

1 1 i l l j . i L

                                                                                                               %        a             *
                                                                                                                 ,00 0      ,

7 i i, (' Natural Draft Cooling Tower Site l CV r .g , * ~2 - (C.ocedval 3 0SNA - ) } Approved Design ateteniel Power Conversion page 10.M7 J

System 80+ Deslan ControlDocument Table 10.4.6-1 Representative Condensate Cleanup SystenhDesign and Operating LTarameters

d Desesiption ParameterN Number oflead cations beds 9 active plus I spare Number of mixed beds 9 active plus 1 spare Diameter of cation and mixed beds (ft) 10.5 Depth of cation and mixed beds (ft) 4 Total volume of cation bed resin 280 Total volume of mixed bed resin (cu ft) 275 Active cation bed regeneration figwsy (days) 8-10 Active mixed bed regeneration frequency (days) 60-90 Total volume of waste water from lead cation bed per regeneration (gal) 30,000 Total volume of waste water from mixed bed per regeneration (gal) 62,000 Number of 100% capacity waste neutralization tanks 2 Gross volume of each waste neutralization tank (gal) 115,000 ,

Max flow rate of neutralization system effluent discharge (gpm) 250ja JN i

                                                                                                                                    . 6 :( ~

('I De Condensate Cleanup System parameters are representative of typical system design and operating conditions. Ac Condensate Cleanup System equipment, regeneration requirements, etc., will be cdjusted based on site specific requirements. The system / component parameters listed in this table support the bounding liquid waste discharge analysis as discussed in Section 11.2.7.1.

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System 80+ Design ControlDocument Adequate space is provided in the Radwaste Building to collect, store, a!xi process radioactive liquid waste, as well as perform maintenance and inspection activities. ' l-

2. A quality assurance (QA) program will be applied with the provisions as specified in regulatory position C.6 of Regulatory Guide 1.143.

l -

3. The LWMS is constructed in accordance with regulatory position C.4.3 by the provision of the following features.

Pressure-retaining components of process systems should use welded construction to the maximum practicable extent. Process systems include the first root valve on sample and instrument lines. Flanged joints or suitable rapid-disconnect fitting should be used only where maintenance or operational requirements clearly indicate that such construction is i preferable. . Screwed connections in which threads provide the only seal should not be used except for instrumentation and cast pump body drain vent connections. Process lines should not be less than 3/4 inches (nominal). Screwed connections backed up by seal welding, mechanical joints, or socket welding may be used on lines 3/4 inches or , larger but less than 2-1/2 inches. For lines 2-1/2 inches and above, pipe welds should be of the butt joint type. Nonconsumable backing rings should not be used in lines carrying resins or other particulate material. All welding constituting the pressure boundr.ry or pressure-retaining components should be performed in accordance with ASME Boiler and Pressure Vessel Code Section IX.

4. The LWMS will be hydrostatically and pressure tested in accordance with regulatory position C.4.4. Testing of piping systems during the Operation phase should be l

performed in accordance with applicable ASME or ANSI codes or system piping ; specifications. For System 80+ repairs, replacements, and modifications less than 1 inch l l NPS are exempt from pressure testing provided the original system was pressure tested. ) l 5. The LWMS is designed to permit periodic testing of active components to evaluate the l operability of the LWMS in accordance with regulatory position C.4.5. I e The Radwaste Building housing the LWMS is designed to meet requirements specified in regulatory positions C.5 and C.1.1.3 of Regulatory Guide 1.143. The Radwaste Building is designed as a Seismic Category II building; however, it is designed for the Safe Shutdown Earthquake (SSE) using Seismic Category I criteria. The Radwaste Building structural design features are discussed in Section 3.8.4.1.7. The Radwaste Building is designed for control motions described in Section 3.7. The Operating Basis Earthquake (OBE) is not in the design basis for the System 80+. Although the likelihood of the OBE is greater than the SSE,' the leads associated with the SSE are higher and govern the design of the plant. The structural design of the Radwaste Building meets Regulatory Guide 1.143 requirem:nts.

The Quality Assurance (QA) program for the design, installation, procurement, and fabrication of LWMS components complies with regulatory position C.6 of Regulatory Guide 1.143. Table 3.2-1gidentifies seismic category, quality and safety class for each of the respective components l in the LWMS-in accordance with a constructors QA program in compliance with federal regulations. ((The COL Applicant will develop a construction and operations QA program.))' p

                       '          COL information item, see DCD Introduction Section 3.2.
                     .%..J Deekre A0eteniel- Wesee ?% -               ^

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Svatem 80+ Desirr Centrol Drcument 11.2.2.2.10 Laundry and Hot Shower Tank The laundry, hot shower and chemical waste subsystem is designed to provide the capability to terminate I the discharge and divert flow to the collection tanks upon detection of high radiation in the discharge. After sampling the detergent waste sample tank contents, the operator would either recirculate liquid I waste to the subsystem collection tank for reprocessing, or manually divert flow to the low level radwaste subsystem for processing, as necessary, based on sampling results. Similarly, the condensate cooler tank i discharge would be automatically termmated upon receipt of a high radiation signal. Subsequently, the J operator could manually divert flow, as ====*ry based on sampling results of the subject tank, for j further processing in the low level waste subsystem. l 11.2.2.3 System Operation During normal operation, each pair of Waste Collection Tanks will have one available to accept waste inputs and the other will be available for processing if necessary. Since the LWMS operators will have level indication on waste volumes, they will anticipate system collection and processing requirements. ; Liquid waste processing and release are separate batched processes. After a Waste Collection Tank has received as much waste as the operators deem appropriate, its inlet valve is closed to permit sampling. i I Generally, the Collection Tank contents are recirculated and mixed when waste liquids are entering the tank and the tank level exceeds a minimum level. This expedites the batch sampling process and assures the final sample will be representative. Based on the results obtained from the initial collection tank sample, appropriate pre-process chemical addition and processing is performed in the Collection Tank, as necessary. Based on final sample results, the decision will be made to process the tank using the existing process vessel or to provide a more appropriate process. Normally, the effluent for the Chemical Waste Tanks and the Laundry and Hot Shower Tanks are not processed and directly discharged. Because of segregation of inputs, the size of the collection tanks, and the flexibility of the normal ion-exchange process, a revised process should not be necessary. However, if a change is considered necessary, it will be implemented based on status of individual process vessels inferred from previous influent and effluent sampling. Re-alignment of the flow path can be rapidly accomplished using remote operated valves. The LWMS subsystems collection tanks are sampled prior to processing. Each respective process stream is processed, as appropriate, based on the sampling results. The contents of the neutralization tank and each of the respective LWMS collection tanks, waste monitor or sample tanks are processed, as

       - necessary, to ensure compliance with 10 CFR 20, Appendix B of Sections 20.1001 through 20.2402, Table 2, Column 2 effluent concentration limits. After processing, the waste water is collected in waste monitor or sample tanks where it is sampled prior to discharge to the environment. A radiation monitor is provided downstream of the last possible input of radioactive liquid waste. Upon detection of a                           :

radiation signal above the monitor setpoint, the discharge would be automatically terminated. ((The g setpoint is determined by the COL Applicant and provided in the Offsite Dose Calculation Manual.))' FT l Section 11.5 provides a more detailed discussion regarding the determination of this setpoint. The j operator would then sample the appropriate tanks in the LWMS subsystem (s) and the neutralization tank l that were being discharged. Based on the sampling results, the operator would manually divert flow, as i 1 i necessary, for further processing as illustrated in Figures 11.2-1, sheets 1 through 4, and Figure 10.4.6-1. It is not expected that re-alignment of the process vessels will normally coincide with changes in collection tanks being processed. This is because, as indicated above, the variation in constituents from COL information item; see DCD Introduction Section 3.2. 4 proved Design hinterial Weste nianagement Pope 11.210

l Syntem 80+ Design ControlDocument l l i l ((The setpoint for the liquid waste management system discharge radiation monitor is determined by the COL Applicant and provided in the Offsite Dose Calculation Manual to ensure compliance with 10 CFR 20, Appendix B of Sections 20.1001 through 20.2402 effluent concentrations.))' -! The radioactive liquid effluent monitor is located downstream of the last possible point of input , of radioactive liquid waste. j High activity is alarmed in the Radwaste Building control room and in the Main Control Room via the Data Processing System (DPS) and the Discrete Indication and ALARM System (DIAS). j l ((The COL Applicant will provide the operational setpoint for the termination of the liquid waste COb-management system contamment cooler condensate tank, turbine building drain, steam generator ! blowdown, and neutralization tank discharges to the environment in the plant-specific offsite dose diJ- , calculation manual (ODCM).))' This setpoint ensures that the ratio of instantaneous concentrations of radionuclides in the liquid effluent in unrestricted areas to the corresponding values given in 10 CFR 20, Appendix B of Section 20.1001-20.2402, Table 2, Column 2 summed over the radionuclides in the liquid effluent does not exceed 10. Prior to release, the regenerant waste water is held in one of the neutralization tanks from which a representative sample is taken. Releases from the neutralization tanks will be batch releases. ! l All releases are made through a process radiation monitor. Upon detection of a radiation signal { above the radiation monitor setpoint, the release is terminated automatically. The contents of the j neutralization tank would then be sampled by the operator and the flow manually diverted to the Floor Drain Tank in the low level waste subsystem of the LWMS for further processing.

Differential Pressure Both Filter and Media Bed Process Vessels are equipped with differential pressure measurement instrumentation to monitor the loading of the filter or bed media. Differential pressure indication is provided in the Radwaste Building control room.

I e Flow

Each Process Pump is equipped with flow measurement to assist the operators in regulating the l l process within the appropriate operating range. Flow rate information, in conjunction with

! differential pressure information, is also important for the operator to assess filter media l condition. Flow indication is provided in the Radwaste Building control room. e Area Radiation 1

Area radiation monitors are discussed in Section 11.5.1.2.5. Area monitors willhave local visual l j and audible alarms.

11.2.6 Estimated Liquid Releases The estimated quantity of radioactivity released in liquid effluents during normal operation, including : operational occurrences, is shown in Table 11.2-1.

                       '        COL information item; see DCD Introduction Section 3.2.

l l Approner outen unserw. wuse unnerement reue 11.212

System 80+ Deslan ControlD:cument The methodology of NUREG-0017 (Reference 1) is used in determining liquid radioactive releases. The sources, estimated volumes, and activity levels of LWMS waste input streams as well as other NUREG-0017 model input parameters and assumptions are summarized in Tables 11.2-2 and 11.2-3. A simplified liquid pathway release assessment process model is provided in Figure 11.2-2. 11.2.6.1 Release Points All discharges from the LWMS subsystems of detectable radioactivity are made through a common discharge header. The LWMS is designed with the capability to simultaneously discharge any or all of the radioactive liquid waste water from the LWMS subsystems' collection and/or waste monitor tanks and the condensate cleanup system neutralization tanks, as appropriate, through a single dedicated discharge point. ((The setpoints on each of the discharge lines will be determmed and coordinated by the COL Applicant, as discussed in Section 11.5. The determination of the setpoints of the LWMS discharge gh radiation monitor, located downstream of the last possible point of input of radioactive liquid effluent gl' from the respective LWMS collection or waste monitor tanks and radiation monitor located downstream of the condensate cleanup system neutralization tanks discharge, will be provided by the COL Applicant. The COL Applicant will develop the setpoints for radiation monitors on each of the discharge lines at the common plant discharge header for radioactive liquid effluents.))' Development of these setpoints is discussed in Section 11.5. All releases are monitored prior to dilution and discharge. Complete mixing of liquid waste with the dilution flow prior to discharge is assured by combining the two flows well upstream of the respective discharge point.

                                                                                                           $rtS.

11.2.6.2 Dilution Factors >

                                                                                                            'Ad The dedicated liquid waste dilution flow can vary depending on the number of Liquid Waste Dilution Pumps that are operating. For the purpose of dose evaluations, an average dilution of 100 CFS is 4J J,u assumed for all release points for potentially radioactive liquid effluent. The 10 CFR 50, Appendix 1       a analysis for the liquid pathways is based on a dilution flow of 100 cfs. ((This dilution flow may be        e,A comprised of dilution flow provided by the following sources as determined by the COL Applicant:))'

4,q

Dilution pumps, Co l, 4L k

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cooling tower blowdown, and/or I
site specific dilution flow parameters (e.g., site specific hydrology).

((The discharge point is assumed to be located on a receiving water such that no significant recirculation M~ occurs between the dilution flow intake and discharge.))2 g,ygggf4 ,,A,

                                                                                          'DOW             CDL n The rate of radioactive liquid discharges will be based on the available dilution and concentrations of 10  %)t u CFR 20, Sections 20.1001-20.2402, Appendix B, Table 2, Column 2.

4 .f-{,, COL information item; see DCD Introduction Section 3.2. l 2 Conceptual Design Information; see DCD Introduction Section 3.4. l Approved Design Mosend Weste Management Page 11.213

Syntem 80+ Deslan ControlDocument ((The COL Applicant will provide the operational setpoint for the termination of the gaseous waste management system, high and low purge, and fuel building ventilation system discharges dL to the environment in the plant-specific offsite dose calculation manual (ODCM).))' This setpoint is based on the instantaneous dose rates in unrestricted areas due to the release of

                                                                                                                             // t/

radioactive materials released via gaseous effluent. This setpoint ensures that the instantaneous dose rates offsite are less than the following: Nobles Gases 500 mrem /yr total body; j 3000 mrem /yr skin Other Gases P 1500 mrem /yr to any organ

Accidental releases of radioactive materials from a single component of the GWMS must not result in offsite doses which exceed the guidelines of Branch Technical Position ESTB-11-5.

Section 11.3.7 provides a discussion of the analysis of a single component failure of the GWMS. The methodology used in this analysis is in accordance with Branch Technical Position (BTP) ESTB-11-5 for the design basis source term. The results of this analysis confirm that the dose consequence of a single failure of a GWMS component is within the dose limits of the Branch Technical Position (500 mrem total body). e The system must also contribute to meeting the occupational exposure design objective by keeping operation and maintenance exposure ALARA. The GWMS is designed in accordance with guidance provided in Regulatory Guide 8.8, ANSI /ANS-55.4, and Regulatory Guide 1.143 and 1.140. This ensures that the GWMS will meet ALARA objectives.

Protection will be provided to gaseous waste handling and treatment systems from the effects of l an explosive mixture of hydrogen and oxygen in accordance with 10 CFR 50, Appendix A (General Design Criterion 3).

i The GWMS is designed to preclude the buildup of an explosive mixture of hydrogen and oxygen

in accordance with the Standard Review Plan, Section 11.3. The charcoal vessels, condenser cooler, piping, analyzer pressure boundary and valves within the GWMS will be designed to withstand a hydrogen explosion (i.e., twenty times normal operating pressure) in accordance with i ANSI Standard 55.4. One hydrogen and one oxygen gas analyzer is utilized to monitor H2 and O2gas concentrations in the GWMS. Alarms are provided locally in the Nuclear Annex and in the Main Control Room to alarm on high oxygen concentration.

11.3.1.2 Codes and Standards The GWMS is designed in accordance with the guidance of Regulatory Guide 1.143 from applicable regulatory positions (C.2, C.4, C.5 and C.6). These include: e The GWMS is designed and tested in accordance with regulatory position C.2 of Regulatory l Guide 1.143. l l t ' 8 COL information item; see DCD Introduction Section 3.2. Swd Desion Meaww. ware uenegement rege y1.3 2 L l .

l i Sv-tem 80+ oeskn contrat Document

4. The GWMS will be hydrostatically and pressure tested in accordance with regulatory position C.4.4. Testing of piping systerns during the Operation phase should be performed in accordance with applicable ASME or ANSI codes or system piping specifications. For System 80+ repairs, replacements, and modifications less than 1 inch NPS are exempt from pressure testing provided the original system was pressure tested.

5. The GWMS is designed to permit periodic testing of active components to evaluate the operability of the GWMS in accordance with regulatory position C.4.5.

The Nuclear Annex housing the GWMS is designed to meet the requirements of regulatory position C.5 of Regulatory Guide 1.143. . The Nuclear Annex is designed to withstand a Safe Shutdown Earthquake (SSE). Supports for the charcoal adsorbers beds are designed to withstand a SSE to meet requirements specified in regulatory positions C.5.1.1 through C.S.I.3. The Nuclear Annex structural design is discussed in Section 3.8.4. The Nuclear Annex is designed for control motions described in Section 3.7. The Operating Basis Earthquake (OBE) is not in ;

the design bases for the System 80+. Although the likelihood of the OBE is greater than the SSE, the loads associated with the SSE are higher and govern the design of the pht. The structural design of the Nuclear Annex meets Regulatory Guide 1.143 requirements

The Nuclear Annex is designed as a seismic Category I building and is designed to withstand a plant Safe Shutdown Earthquake (SSE) and meets the seismic requirements specified in regulatory position C.S.
The Quality Assurance (QA) program for the design, installation, procurement, and fabrication of GWMS components complies with regulatory position C.6 of Regulatory Guide 1.143. Table l 3.2-1 identifies seismic category, quality and safety class for each of the respective components l

in the GWMS in accordance with a constructors QA program in compliance with federal Co L regulations. ((The COL Applicant will develop a construction and operations QA program.))' f7-/ 11.3.1.3 Features l The following features assist in meeting the Design Criteria:

High activity hydrogenated gaseous waste streams are processed through carbon adsorbers to retain and delay radioactive fission gases prior to release. Decay of fission product gases (xenon and krypton) prior to release significantly reduces offsite exposure levels. Additionally, filtration of process ventilation systems significantly reduces offsite exposure from radioactive particulates and iodine in plant effluents.
The carbon adsorber delay process used is simple with minimal active components. Features to facilitate cleaning and/or otherwise improve radwaste operations are provided. Additionally, a spare carbon bed is provided in the event an operating bed becomes inoperable and provisions for nitrogen dryMg of moisture contaminated beds are provided.

    '       COL information item; see DCD Introduction Section 3.2.

Approved Design Meterial Weste Menegement Page 11.34

~~

Sv tem 80 + Design ControlDocument ; I o The Radwaste Building housing the SWMS is designed to meet the requirements specified in ' regulatory positions C.5 and C.1.1.3 of Regulatory Guide 1.143. The Radwaste Building is designed as a Seismic Category II building; however, it is designed for the Safe Shutdown Earthquake (SSE) using Seismic Category I criteria. The Radwaste Building structural design features are discussed in Section 3.8.4.1.7. The Radwaste Building is designed for control motions described in Section 3.7. The Operating Basis Earthquake is nct in the design basis for the System 80+. Although the likelihood of the OBE is greater than the SSE, the loads associated with the SSE are higher and govern the design of the plant. The structural design of the Radwaste Building meets Regulatory Guide 1.143 requirements, o The Quality Assurance (QA) program for the design, installation, procurement, and fabrication of SWMS components complies with regulatory position C.6 of Regulatory Guide 1.143. Table 3.2-1 identifies seismic category, quality and safety class for each of the respective components in the SWMS in accordance with a constructors QA program in compliance with federal g regulations. ((The COL Applicant will develop a construction and operations QA program.))' ,pj 11.4.1.3 Features The following features assist in meeting the design criteria. I o The system has provisions to accommodate leased equipment which may provide the most economical choice at particular times or for particular waste. o Normal system operations are remotely controlled from the Radwaste Building Control room l which permits operators to most effectively coordinate activities. o Active and replaceable components have crane or monorail hoist access to facilitate removal and j repair. 11.4.2 System Description l 11.4.2.1 General Description Primary functions of the SWMS include providing means by which spent resin, filters, etc. from the l LWMS and primary letdown systems are processed to ensure economical packaging within regulatory guidelines, as well as handling dry, low activity wastes for shipment to a licensed burial facility. The spent resin transfer system is designed to transfer expended radioactive demineralizer and ion exchanger resins from their vessels to the spent resin tank. The system also provides holdup of the resin l and transfer of the resin to the solidification system. The major components of this system are spent resin l tanks, spent resin surge tanks, spent resin transfer pumps, and filters. The spent resin transfer system is shown on Figure 11.4-1 (Sheet 1 and 2). The spent resin tanks provide settling capacity for radioactive bead resins transferred from various demineralizers. Capability is provided for solidification of dewatered resins or sluicing to containers approved for shipping and disposal of dewatered ion exchange resins. Also, connections are provided for use of vendor supplied services such as rapid dewatering or waste drying systems when it is

         ' COL Information item; see DCD Introduction Section 3.2.

4Rprovmf Doelptr Metenlof Weste Menegemorrt Page 11.4 5

System 80+ D sign Contrri Document If necessary, low activity resin is sluiced to the low activity spent resin tank to allow for settling and holdup prior to processing. Otherwise, they are batched directly to disposal containers for vendor-service processing and direct shipment to a licensed burial facility. High activity spent resin is sluiced to the high activity spent resin tank to allow settling and decay of short-lived isotopes. Resin is then transferred to the dewatered waste processing High Integrity Containers (HICs) located in the Radwaste Building. In some cases, high activity resins may be blended with low activity resins. Blending may be accomplished by utilizing a cross connection which allows transfer of low activity spent resins to the high activity spent resin storage tank.

?or the purpose of the radioactive liquid waste discharge analysis, discussed in Section 11.2, it has been assumed that all resins from the condensate cleanup system polishers will be regenerated. The condensate cleanup system is discussed in Section 10.4.6. As necessary, the resins from these polishers will be processed and packaged for disposal if the resins become physically broken or the decontamination efficiency is reduced. At this time, the resins will be collected in the spent resin decanting tank located in the Turbine Building. This tank is a process tank, not a storage tank. The spent resin decanting tank will be sampled and if the resins are radioactive, they will be sluiced into the shipping container and dewatered in the Turbine Building. A high integrity container, HIC, will be used only as necessary to ensure compliance with state, federal, and Department of Transportation regulations. After the spent resins are dewatered, the shipping container (e.g., HIC or drum) will be transported to the Radwaste Building for interim storage until shipment to a licensed burial facility.

If the spent resins are non-radioactive, the resins will be disposed of in accordance with state and federal regulations. The resins may be categorized as chemical or industrial wastes. ((The COL Applicant will ff 2-address categorization and disposal of this waste.))' The blowdown demineralizer resins will also be processed in the Turbine Building similar to the condensate cleanup system resins. However, the blowdown demineralizer resins will not be regenerated. Decant from the HICs, removed during the dewatering process, is directed back to the LWMS for processing prior to release to the environment. Non-clogging screens on the spent resin tank and filters in the process line are provided to prevent the carryover of spent resin beads or fines to the LWMS during the transfer of decanted water. When solidification of spent resins is desired, instrumentation on the spent resin storage tank is used to assure that the appropriate water-to-resin ratio is present. Adjustments to the ratio may be made using available water supplies or the spent resin forwarding pump as necessary. Following mixing, valve alignments are made to the binding area. Filled containers (such as HIC, or drums) may be stored in the shielded storage area until shipped. 11.4.2.3.2 Spent Filter Storage and Handling An area is provided in the Radwaste Building for storage of the process and HVAC filters used throughout the station. The services are summarized as follows. ' COL information item; see DCD Introduction Section 3.2. Approved Design Material- Weste Management Page 11.4-9

Syntem 80+ Design ControlDocumen{

The System 80+ design of the Process and Effluent Radiological Monitoring and Sampling Systems provides instrumentation to measure, record, and readout in the Main Control room, as well as control releases of radioactive materials in plant process systems and effluent streams.

This system is designed to provide for continuous sampling and monitoring of radioactive iodine and particulate, as well as the capability to take grab samples in gaseous process and/or effluent streams in all potential accident release points. A particulate / iodine fixed filter canridge is provided for all plant ventilation systems, with the exception of the nuclear annex and radwaste building ventilation system, which have their own particulate and iodine monitoring systems. Except for the turbine building exhaust, containment purge, the main condenser evacuation system, the Nuclear Island ventilation systems, and the Gaseous Waste Management System, exhausts discharge through the unit vent. Provisions for taking grab samples are provided as specified in Table 11.5-6. Additional discussion regarding sampling capabilities for gatcous process and effluent streams is addressed in Section 11.5.2.2. A fixed iodine absorption filter arai detector assembly, as well as a moving filter and detector l assembly are provided for the unit vent monitor as discussed in Section 11.5.1.2.3.1. The ! ventilation systems are provided with a fixed iodine absorption filter and detector assembly only, l with the exception of the Nucl:ar Annex and Radwaste Building ventilation systems which are provided with its own paniculate and iodine detection systems discussed in Section 11.5.1.2.4. The capability for taking grab samples from the unit vent and ventilation system exhausts are oC provided, as specified in Table 11.5-6 at the respective radiation monitor locations. ((These grab p$ 1 samples are taken for aralysis, at a frequency established by the COL Applicant))"I, onsite to the primary chemistry lab and counting room during normal operating and post-accident conditions. , 1 i Continuous sampling of all potential post-accident release points is provided by the continuous unit vent sampler. This sampler contains a fixed filter particulate and iodine cartridge which reeives continuous flow of sample air from the unit vent duct. This sample canridge is routinely replaced and taken for detailed onsite lab aratory analysis in the primary chemistry laboratory and counting room, where precise assessment of releases is performed for the period during which the canridge collected the sample. ((The frequency at which the canridge is replaced and (d analyzed is determined by the COL Applicant and specified in the operations and mairaenance //-5 , manual.))UI The sampler is designed to be used during normal operation and post-accident l conditions to meet the sampling requirements specified in 10 CFR 50.34 (f) (2) (xvii) and l NUREG-0737, Attachment 2, Section II.F.1. ! The Process and Effluent Monitoring and Sampling System is designed with a continua.is control room interface via the DPS and DIAS systems. Primary indication of radiation levels and status of alarms for post-sccident and non-post-accident radiation monitors are processed through the DPS and DIAS systems. , ((The COL applicant will demonstrate conformance with 10 CFR 50 Appendix I, ANSI N13.1, RG 1.21 I and RG 4.15.))"I l l l Ul COL information item: see DCD latroduction Section 3.2. Approved Design Material Radiation hotection Page 11.5-2

System 80+ Deskn ControlDocument Secro. n> // S l.\ Area radiation monitors are also used in special process applications. Area radiation monitors located next to the main steam lines are used for monitoring a steam generator tube leak or rupture and high range area detectors are used to estimate post-accident containment airborne activity and primary coolant activity. A high range area detector is also located near the reactor coolant purification filters to monitor the accumulation of activity on the filters to prevent the radiation level from exceeding the shielding capacity of the spent filter transfer cask and to support solid waste disposal. d.- 0- A M The ranges and sensitivities of the monit s are based upon th maximunyand/ minimum expected concentrations for normal plant operation, including anticipated ransients,8 and 0 postulated accidents in accordance with 10 CFR 20 limiis and regulatory guidance. ((The range and sensitivity values listed in Tables 11.5-1 through 11.5-5 represe ' isign values which meet or exceed the regulatory guidance and are generally commercially availab .))y 11.5.1.2 System Description 11.5.1.2.1. Monitor Design and Configuration Process and effluent, and airborne radiation monitors typically consist of components such as a microprocessor, one or more detectors, a shielded detection chamber, a sample pump, flow instrumentation, and associated tubing and cabling. Three basic types of process and effluent monitoring configurations are used: off-line, on-line, and in-line. In the off-line type system, a representative sample is taken from the fluid and routed through a filter or chamber for detection of activity. Off-line type systems allow optimized design of detector geometry and shielding, and for detection chambers to be located away from areas of high background radiation. Most l off-line monitors are designed as factory-built package skids; however, some applications have a l microprocessor which is mounted remotely from the detector / shield assembly. While most off-line monitors utilize a sample pump to draw a sample, some off-line monitors use system pressure to create sample flow. The on-line system configuration has the detector located next to or on the pipe, duct, or tubing carrying the process or effluent fluid. This type system has the advantage of being simpler (fewer active l components) and having a quicker response time. These type monitors generally have less shielding than off-line monitors and have remote mounted microprocessors. l l In-line type systems are very similar to the on-line type systems except that the detector is located in the l actual process stream. The detector is generally mounted inside a well which allows the detector to be l surrounded by the process or effluent fluid. i Each process and effluent, and airborne monitor is located in an easily accessible area and is provided ! with sufficient shielding to ensure that the required sensitivity is achieved at the design background radiation level for the area. A checksource is used to periodically check the operability of each detector. The checksource is automatically activated by the menitor microprocessor at pre-set intervals. In addition, manual actuation of the checksource is available in the Main Control Room and at the microprocessor location. Instrumentation and sensors are provided on monitors to detect component failures such as loss of sample flow, clogged or torn filter paper (if applicable), and loss of detector counts. ,j m i l ') , Conceptual Desi G- -. gn Information; ,. . c ~ see DCD Introduction Section 3.4. , , , . ~

i System 80+ Design ControlDocwnent 1 discharge point for potentially radioactive airborne and gaseous effluents from the plant. A l multiprobe isokinetic sampler is used to draw a sample from the Unit Vent at a point downstream of the last point where flow can be added to the unit vent. Special design consideration is given j to sample tube routing to ensure sample integrity of particulates and iodine. A moving filter paper detector assembly is used to continuously scan particulate activity deposited on the filter. A fixed iodine absorption filter and detector assembly is used to collect and monitor iodine in the effluent stream. A Nal detector system with a single channel analyzer is used to look specifically at I-131. Finally, the sample flow is monitored by low and high range gaseous ! activity detectors before being returned to the vent. l The design of this monitor is in compliance with NUREG 0737 and Regulatory Guide 1.97 ! requirements. The location of this monitor is chosen to minimize sample tube length for quicker monitor response and better sample integrity. The monitor is also located away from high post-l accident radiation areas and is designed with sufficient shielding to prevent expected post-accident I radiation from interfering with obtaining the necessary sensitivity. I The monitor and sample tubing system are designed in conjunction with a continuous unit vent sampler. This sampler contains a fixed filter particulate and iodine cartridge which receives a continuous flow of sample air. This sample cartridge is routinely replaced and taken for detailed laboratory analysis where a precise assessment of releases is performed for the period during { which the cartridge collected the sample. The sampler is designed to be used for normal ! operation and post-accident sampling requirements. These sampling requirements are discussed in Section 11.5.2. l e Unit Vent Post-Accident Monitor , I The unit vent post-accident monitor continuously monitors the airborne radiation levels in the unit vent by means of an ion chamber detector located in the well of an in-line sampler. This monitor allows the operator to observe the radioactivity releases which could result from abnormal occurrences or accidents. This type of system is very simple and reliable and provides backup indication to the unit vent monitor. This monitor is used in conjunction with the off-line Unit l Vent Monitor (above) to cover the full range of radiation detection required by Regulatory Guide 1.97. I e Containment High Purge Exhaust and Low Purge Exhaust Monitors l These monitors detect radioactivity in the high purge exhaust and low purge exhaust lines, respectively. On-line monitors are used for this application. High activity indication from either of these monitors automatically terminates the associated containment purge. The containment high purge exhaust monitor, in conjunction with the containment atmosphere radiation monitor, provides the information required for the conduct of the containment purging operation. Sample taps are provided on both the High and Low Containment Purge Exhaust lines for taking particulate or iodine grab samples. ((The COL Applicant will verify that the range specified in Table 11.5-1 can be met.))VI l t VI COL information item: see DCD Introduction Section 3.2. A; proved Design Material- Radiation Protection Page 11.5-8 I t

System 80+ Design ContrriDocument Also, sampling of some systems is necessary because of required actions in the Technical Specifications when certain continuous monitors are out of service. ((The Technical Specification actions specify the , frequency of sampling and any other special requirements which apply to the sampling procedure developed by the COL Applicant.))N 11.5.2.3 Expected Composition and Concentrations The specific radionuclide compositions will vary for each batch release from the contaiament or the liquid radioactive waste management system and fluctuate somewhat from day to day from the unit vent continuous releases due to variations in plant operating conditions. The estimated radioactive releases for liquid effluents are given in Table 11.2-1. The estimated annual airborne effluent releases are given in Table 11.3-4. 11.5.2.4 Sampling Equipment and Procedures Samples are collected by plant technicians and analyzed and measured in the counting room in accordance with station operation procedures concerning the release of radioactive waste. The frequency of sampling is in accordance with Regulatory Guide 1.21 guidelines. Station sampling procedures will establish methods of sampling for each sampling location to assure that representative samples are taken and that these methods will be consistent for all personnel performing the sampling. Tables 11.5-6 and 11.5-7 , list the process systems, sampling capabilities, sampling provisions, and the approximate location that samples are taken from gaseous and liquid process and effluent streams, respectively. ((The COL Applicant will determine the sampling frequency, purpose, sensitivity, and type of analysis and provide ,

them in the operations and maintenance manual.))' The liquid contents of a tank being sampled are recirculated prior to taking the sample to ensure thorough mixing of sediments and particulate solids in the tank. All sample connections are located in a free flowing stream or in a location where a representative sample may be taken. The sample lines are purged for an adequate period of time before the sample is taken to ensure that the sample is representative. ; i Effluent ventilation ducts are sampled isokinetically in accordance with ANSI N13.1 for radioactive gases, ; particulates, and iodines. Iodine samples are collected using special iodine filtering cartridges and taken ' to the counting room for analysis. Particulate sampling utilizes fixed paper filters for laboratory analysis. Gas sampling utilizes special gas collection canisters which allow easy connection and disconnection from : sample taps for transport to the counting room. ' 11.5.2.5 Analytical Procedures and Sensitivity Samples of process and effluent gases and liquids are analyzed in the counting room in accordance with station procedures and Regulatory Guide 1.21. Analytical procedures used are based on methodology utilized in general practice in the nuclear industry or in applicable standards and the accuracy and precision of the results are standardized with central or outside laboratories using radioactivity standards traceable to the National Bureau of Standards. Laboratory equipment is provided for the counting room to perform gross beta counting, gross alpha counting, gamma spectrometry, liquid scintillation counting, and radiochemical separations. N COL information item; see DCD Introduction Section 3.2. Approved Design Material- Radiation Protection Page 11.5-ti I l 1

Systzm 80+ Design Contr:IDxcument l j Table 11.5-4 Area Radiation Monitors

Monitorm Typical Range Power Source Seismic (mR/hr) Catego Reactor Containment Entrance 0.1 - IE+7 non-1E None Refueling Bridge Crane 0.1 - IE+4 non-lE 11 In-core Instrumentation Equipment 0.1 - IE+4 non-1E None Decontamination Area 0.1 - IE+4 non-lE None Sample Room 0.1 - IE+4 non-lE None Main Control Room 0.1-IE+4 non-lE None Primary Chemistry Laboratory 0.1 - 1E+4 non-1E None New Fuel Storage Area (2) 0.1;1E+4 non-lE None Spent Fuel Pool Bridge 0.1 - IE+4 non-lE 11 Fuel Building Area 0.1 - IE+7 non-lE None Nuclear Annexm (normal operation) 0.1 - IE+4 non-lE None Nuclear Annexa (post-accident) 100 - IE+7 non-lE None Solid Waste Drum Storage and Handling 0.1 - IE+4 non-lE None Area Radwaste Building leading Bay 0.1 - IE+4 non-lE None Hot Machine Shop 0.1-IE+4 non-lE None Hot instrument Shop 0.1-IE+4 non-lE None l Radwaste Building Areasm 0.1-IE+4 non-lE None l Reactor Building Subspherem (normal 0.1-1E+4 non-lE None operation) Reactor Building Subsphere (post. 100 - IE+7 non-1E None accident)A Technical Support Center Area 0.1 - I E +4 non-lE None i A The preliminary locations for area radiation monitors are provided in Figures 12.3-9 through 12.3-16, and 12.3-18 through 12.3-21. cp b A hlecations of Post-Accident Area Monitors in the Nuclear Annex and Reactor Subsphere are t-based on Y (V accident shielding analysis and access study which will be provided later. A The initial design allows for a number of additional channels to be added to the RMS for the Nuclear Annex, Reactor Subsphere, or Radwaste Buildings during the final design or after start-up as is determined ! necessary for personnel protection. Final selection oflocations will be based on final equipment layout and i [41 system tm: ,oa operating characterNtics. eoL aform>hin kb rokeluck Seebk 3 2.. Approved Design Meterie!

Radiation Proseceien Page 11.5-24

System 80+ Design Cenerof Docenerrt e Protects certain components from excessive activation or excessive radiation exposure. e Facilitates access for maintenance of components. Maimmining occupational exposures as low as reasonably achievable (ALARA) is a major design , 4 consideration in accordance with Section C.1 of Regulatory Guide 8.8. Although the station's design is ; intended to result in ALARA exposures during operation, these design features will also facilitate ' M'wnmiazioning, Regulatory Guide 8.8 also provides guidance for both equipment selection and plant layout, hT-selection for System 80+ plays an integral part for maintaining personnel exposure ALARA. Criteria used for equipment selection are discussed in written ALARA guidelines. System 80+ design includes: e Enhanced reliability of equipment which reduces the frequency of mamtenance and the personnel exposure associated with the mamtenance. This is illustrated by the following:

1. Use of reliable extended service lamping in high radiation areas, as well as, location of lighting fixtures so that mamtenance can be performed in a lower radiation area. These features are in accordance with Regulatory Guide 8.8 Position C.2.i.

fro 2. Use of ion exchangers instead of evaporators except in the Chemical Volume and Control y System. The System 80+ design will minimize the use of evaporators based on industry j g experience. Ion exchangers are simpler in design and are more reliable than evaporators. ! 14 e Careful attention to environmental qualification of equipment. This includes a=4nmaar qualification for a variety of environmental conditions, such as radiation, humidity, and temperature. This is discussed in greater detail in Section 3.11. ^- go e Electrical components, containing radiation-sensitive materials, will be shielded or located in low-gh5TN g N radiation . % COL applicant will establish environmental qualification criteria for 89WPmentJ f g o Material selection of piping and components, such as valve seats, in the primary system. tJFLC_ Material is selected with low cobalt or nickel impurities. This minimizes the production of e corrosion products which is a significant contributor to personnel exposure received during mamtenance and operational activities. y have o adle.1 b Maintainahility which includes plant layout and equipment spacing.

                                                                                                                                         ^%h.

The plant layout is designed to maintain personnel exposures ALARA during normal and post-accident conditions. For example, the plant layout is designed to provide: p tt 1. equate spacing to facilitate accessibility of equipment during maintenarse activities, 2. [he t spread of contamination,paration of non-radioactive syste i 00C ihedh4 ~kwi, we 1cb Thd.icMirn E.hu 3,2 4pment oseen asesenw Aseseehn preescalen pape ra.r.3

System 80+ Deefan ContralDocument Personnel and job exposures trends are reviewed by management at the site and offsite radiation protection personnel, and appropriate action is taken. Summary reports of occupational exposure are provided that describe problem areas where high radiation doses are encountered and that identify which work group is accumulating the highest doses. Recommendations are then made for changes in operating, maintenance, and inspection procedures or for modifications to the station as appropriate to reduce dose. From industry experience it has been shown that the majority of exposure at operating plants is received durmg plant outages from maintenance and inspection activities and not from nonnal operating activities. This is logical since operators can nonnally stay outside shield walls to read instnanents or operate valves and have to r cubicles containing radioactive equipment for short periods of time only to check . equip maintenance and inspection personnel usually must go inside cubicles or behind [ shield w s must be in close proximity to the lines, valves, instruments, or other pieces of equipment ! which are radiation sources. The System 80+ design incorporates lessons learned from past designs, as well as the recommendations from Regulatory Guide 8.8 and 8.10. Design features such as:

1. Addition of platforms around the steam generator and the reactor coolant pump seal ,

cartridges,

2. rovisions for sufficient spacing for equipment laydown/ pull areas,

3. of removable insulation,

4. [ eduction of length of welds by use of seamless piping, and ;

5. [tegrated reactor head removal g

                        %k ensure, personnel exposure is maintained ALARA during maintenance and inspection activities. System g           80+ design features are discussed in further detail in Section 12.3 and the System 80+ ALARA
 <I h. Guidelines Manual.

G g Areas which house systems and components subject to in-service inspections that are high radia are including designedlaydown areas, tobetween permit prompt equipment ingress to facilitate and and maintenance egress. inspectionSystem 80+ is also des activities. Radiation p Protection personnel will perform surveys of areas requiring access for maintenance and inspection

  *y        activities prior to entry and will perform periodic inspections during work activitiesdThe COL A                                       01N will provide Radiation Protection Procedures to provide survey requiremen                                                               gd V

in addition, System 80+ provides the following design features to maintain personnel exposure ALARA E ' during ISI: ( I e Pipe stops, snubbers, and pipe hangers near welds that require ISI or repair are carefully *' positioned to facilitate weld accessibility, e Integrally forged components and seamless pipe is selected whenever possible to avoid in-service weld inspections. This reduces the length of pipe requiring in-service weld inspections.

              .i. COL                       n'rfp rmb%              LW K-G =' Dee4pn aseewint - nmerenen honecanon any, r2.r.s

i l CESSARRBW,m s 1 I / Activity source term for room ( Ci-ft 3/ml-hr) Q m Q = L PFxAo 1_ , l- i ' L = Leakage or evaporation rate (ft /hr liquid phase 3 basis) PF a Partition Factor (The ratio of the quantity of a - nuclide in the gas phase to the total quantity in both the liquid and gas phases when the liquid and - the gas are at equilibrium.) l Ao a Initial activity of fluid stream source (pCi/ml) j F u Roon exhaust flow rate (ft /3 min) h 60 m Conversion factor (min /hr) Credit for decay has been neglected for conservatism. The I airborne concentrations in

               ] personnel throughout the plant will be na      rooms        or   cubicles        accessible      to ned vi in inplant concentrations prescribed in 10 CFR 20.                          Th desi
        ..           criteria neces         to     sureplantdesignlcompl                            e acceptance with  10 CFR g ,K

' 20 requirements a ified ip the Radiation action Design Acceptance criteri document. The design a opta i specified include. e criteria t !

a. Maintai inplant airborne

                                                                                                                     $tu -

concentrations 'of radioisotopes within 10 CFR 20, Appendix B of Sections : 20.1001-20.2402, Table 1, il 1 requiring infrequent access,Column 3 limits for areas 6MokldwT j b. Maintain inplant airborne concentrations ' of radioisotopes well within or a small fraction (i.e. , t 0.25) of 10 CFR 20, Appendix B of Sections Ut 20.1001-20.2402, Table 1, Column 3 limits for areas requiring continuous or frequent access. Inplant airborne in Section concentrations 12.2.2. will be calculated as described

c. Provide

                                                                                                                        @h sufficient              containment and ventilation                        i capability       to prevent                                                                '

contamination. the spread of airborne -

d. Provide airborne radiation monitors in areas normally occupied and where a potential of airborne contamination exists. The airborne radiation ;

monitoring system is discussed in Section 11.5.

      ^      I The COL Applicant will evaluate the plant design in accordance                                     7 f

with this Design Acceptance Criteria document to assure airborne l

              - concentrations are maintained less than 10 CFR 20.1204 limits.

Amendment U 12.2-14 December 31, 1993

1 )

Systzm 80+ Design ContnlDocument

!d 13.0 Conduct of Operations t i 13.1 Organizational Structure of Site Operator

                                                                 &fek Pms(pA. @c[Nd (SAO-                                    ;

l {[Information concerning the site operator' ganizational structure is within the site operator's scope and i shall be provided in the site-specific . This will include site-specific information to address the issues j in the TMI Action Plan Items I.A.I.4 and II.J.3.1.] 1-1 l 3 1 I< l l 1 I l j i I 5 ! \ 1 1 l 1 1 i H l 4 e

      /g      b CL i4              (h le ' t W*  ,, { u_ 3 LD T.'. ,t dse D$y 0 f u k's, 3, L ,

j Approwd Design Metenal- Operations Page 13.1-1

Syntem 80+ - Desian controlDocument , 13.2 Training kInformation concerning the site operator's training program is within the site operator's scope and be provided in the site-specific SAR. This will include site-specific infonnation to address the issues in TMI Action Plan Items I.A.4.1(2), I.A.4.2, and II.K.1(26)))1 l

                                                                                                                          .i i

i 1 l I

             ,          h    Ik    .. C
            -            e .- e.~.                                                                            ,    ,,.,.,

System 80+ Design ControlDocument 13.3 Emergency Planning 13.3.1 Preliminary Planning l 1 1 4 See site-specific SAR. l 1 13.3.2 Emergency Planning ((See site-specific SAR.))l 13.3.3 BOP Interfaces I 13.3.3.1 Technical Support Center 13.3.3.1.1 Summary Description The Technical Support Center (TSC) is an onsite facility located adjacent to the control room that I provides the capability for plant management and technical support to the reactor operating personnel located in the control room during emergency conditions. The TSC is located within two minutes walking time of the control room. In addition, the TSC design incorporates a viewing gallery which is located so as to permit visibility into the control room. It has technical data displays and plant records : available to assist in the detailed analysis and diagnosis of abnormal plant conditions and any significant l release of radioactivity to the environment. The TSC is the primary communications center for the plant during an emergency. The TSC assists the control room operators by providing resources and facilities for handling the administrative items, technical evaluations, and contact with offsite activities. Until the TSC is activated, all functions of this facility are performed in the control room. When the TSC is functional, emergency response functions, except direct supervision of reactor operetions and manipulation of reactor system controls, shift to the TSC. Plant administration, technical support functions, and contact with offsite activities to assist the control room operators are performed in the TSC throughout the course of an accident. Table 13.3.3-1 outlines the transfer of emergency response functions from the control room to the TSC under the various emergency classes, j The level of staffing of the TSC may vary according to the severity of the emergency condition. The ! staffing criteria for each emergency class is fully detailed in the licensee's emergency plan. Section l 13.3.3.1.5 defines the TSC work space size. l 13.3.3.1.2 Function The TSC is a licensee controlled and operated support center that provides the following functions: e Provides for plant manarement and technical support to plant operations personnel during , emergency conditions. i i 1 8 COL information item: see DCD Introduction Section 3.2. Approved Design Material- operations Page 13.31

System 80+ Deslan controlDocumart

              '*           Relieves the reactor operators of peripheral duties and communications not directly related to reactor system manipulations during emergency conditions.

Prevents congestion in the control room during emergency and/or normal conditions.
Performs Emergency Operations Facility (EOF) functions for the Alert Emergency class and for the Site Area Emergency class and General Emergency class until the EOF is functional.

The TSC is the emergency operations work area for designated technical, engineering, and senior licensee i management personnel, any other licensee-designated personnel required to provide the needed technical support, and a small staff of NRC personnel. The resources of the TSC are sufficient such that a senior licensee official can provide guidance and technical assistance to the operating supervisor in the control room. However, all manipulations are performed by the control room licensed operators. The TSC has facilities to support the plant management and technical personnel who are assigned there durmg an emergency and is the primary onsite communications center for the plant during the emergency. j The TSC fyilities may be used by designated operating personnel for normal daily operations, as well as for traimng and emergency drills. Use of the TSC facility during normal operation is limited to activities that will not degrade TSC preparedness for reactor abnormal conditions or reduce TSC systems reliability. : The TSC viewing gallery is designed so personnel situated therein will have a clear unobstructed view - of the controlling workspace as well as the IPSO. The gallery will facilitate a better awareness of control . room activities for personnel located within the TSC. In addition, during normal operations, the gallery will function as a visitors area and thus serve to minimize distractions in the control room due to the presence of guests. ! l 13.3.3.1.3 Location > l - To optimize communications, the TSC is located as close as possible to the control room within the control building. The walking time from the TSC to the control room does not exceed 2 minutes. This l close location allows the senior plant manager, normally working in the TSC, to quickly enter the MCR : for face-to-face interaction with control room personnel. This proximity also provides access to i information in the control room that is not available in the TSC data system. Provisions are made for the safe and timely movement of personnei between the TSC and the control ! room under emergency conditions. These provisions include consideration of the effects of direct radiation and airborne radioactivity from in-plant sources on personnel traveling between the two facilities. 13.3.3.1.4 Staffing and Training .

           ,r                                                                          -

l The site-specific SAR addresses all staffm' g and training issues. L- ~ l i l { FM I j w oo u.~w-e.a. e. ,. rs.s.2 l

l Sy ttm 80+ Design CrntrolDocu:nent operating conditions.1The licensee provides a means for TSC telephone access to commercial te'ephone common-carrier services that may be susceptible to loss of power during emergencies. The licensee ensures that sparegommercial telephone lines to the plant are available for use by the TSC during emergencies. ] The TSC voice communications equipment includes:

Hotline telephone (located in the NRC consultation room) on the NRC Emergency Notification System (ENS) to the NRC Operations Center.
Dedicated telephone (located in the NRC consultation room) on the NRC Health Physics Network (HPN).
Dial telephones that provide access to onsite and offsite locations.
Communications to licensee mobile monitoring teams and to State and local operations centers prior to EOF activation.

The TSC communication system also includes designated telephones (in addition to the ENS and HPN telephones) for use by NRC personnel. Facsimile transmission capability between the TSC, the EOF, and the NRC Operations Center is also provided. 13.3.3.1.8 Technical Data and Data System The TSC Technical Data System receives, stores, processes, and displays information acquired from different areas of the plant as needed to perform the TSC function. The data available for display in the TSC is sufficient to enable the plant management, engineering, and technical personnel assigned there to aid the control room operators in handling emergency conditions. The data system provides access to accurate and reliable information sufficient to determine:

Plant steady-state operating conditions prior to the accident.
Transient conditions producing the initiating event.
Plant system dynamic behavior throughout the course of the accident.

The TSC data system can be used for:

Reviewing an accident sequence.
Determining appropriate mitigating actions.
Evaluating the extent of any damage.
Determining piant status during recovery operations.

s- COL & .. J Approved Design Materiel- Operations Page 13.3-4

1 i l l System 80+ Design Control Document l i 13.3 3.1.9 Records Availability The TSC includes provisions for a complete and up-to-date repository of plant records and procedures at the disposal of TSC personnel to aid in their technical analysis and evaluation of emergency conditions. .

13.3 3.2 Emergency Operations Facility 1 [lThe Emergency Operations Fac[y (EOF) is an out of scope item which shall be provided by the license applicant and is site specific]The fc!!owing acctions contain a description of a typical EOF and interface requirements which must be met to ens. re adequacy with the System 80+ Standard Design. The word ; l "shall" is used to distinguish inad_e requirements that are mandatory from text that is purely l ! descriptive. l l 1333.2.1 Summary Description kThe Emergencyppe ions Facility (EOF) is a nearsite suppon facility for the management of over licensee emergency response (including coordination with Federal, State, and local officials), coordinationw 4 %q of radiolo ical and environmental assessments, and determination of recommended public protective %f ' actions] e EOF has appropriate technical data displays and plant records as discussed in the site-specific AR. t)g ! When the EOF is activated, the functjofis of providing overall emergency response management, '", I monitoring and assessing radiologi uent and the environs, making offsite dose projections, providing ! i recommendations to State and local fficials, and coordinating with Federal officials shift to the EOF in l accordance with site procedures. ransfer of emergency response functions from the control room to the i i EOF under the various emergen classes shall be accomplished as outlined in Table 13.3.3-1. The EOF l shall be designed to the habitability criteria given in Table 13.3.3-2. l 13.3 3.2.2 Functions l 1 The EOF is a licensee controlled and operated offsite suppon center. The EOF shall have facilities for:

Management of overall licensee emergency response.
Coordination of radiological and environmental assessment.
Determination of recommended public protective actions.

l

Coordination of emergency response activities with Federal, State, and local agencies.

i When the EOF is activated, it is staffed by licensee, Federal, State, local and other emergency personnel G+ designated by the emergency plan to perform these functions. It is the location where the licensee j I provides overall management of licensee resources in response to an emergency having acnni or potential environmental consequences % o% 4 g Tc- h Facilities shall be provided in the EOF for the acquisition, display, and evaluation of all radiological, meteorological, and plant system data pertinent to determine offsite protective measures. These facilities are used to evaluate the magnitude and effects of actual or potential radioactive releases from the plant b d hhkn hmba, J(c ]cd dng kchq &ch,3 3,v, Approved Design Material- Operations Page 13.3-6 f r.O,h

System 80+ Design ControlDocument 1 and to determine offsite dose projections. Facilities used in performing essential EOF functions shall be l located within the EOF complex; however, supplemental calculations and analytical support of EOF evaluations may be provided from facilities outside the EOF. The licensee also may use the EOF as the l post-accident recovery management center. l [The EOF allows the licensee to coordinate emergency response activities with local, State, and Federai agencies, including the NRC. Licensee personnel in the EOF use the evaluations of offsite effects to j make protective action recommendations for the public to State and local emergency response agencies. ! bo, Il State and local agencies are responsible for implementing emergency response actions involving the A. , general public. The State and local agencies may operate from the EOF or from their own control *1p % centers at other locations, dependent upon the site-specific provisions of the emergency plan at each plant. 13.3.3.2.3 Staffing and Trainmg 4 { The licensee shall address all staffing and training issues.] ~ l 13.3.3.2.4 Size The EOF building or building complex shall t>e large enough to provide the following: e Working space for the personnel assigned to the EOF as specified in the licensee's emergency plan, including State and local agency personnel, at the maximum level of occupancy without crowding (minimum size of working space pmvided is approximately 75 sq ft/prson). l The EOF working space is sized for at least 35 persons, including 25 persons designated by the licensee,9 persons from NRC, and 1 person from FEMA. This minimum size is increased if the maximum staffing levels specified in the licensee's emergency plan, including representatives from State and local agencies, exceed 25 persons. e Space for EOF data system equipment needed to transmit data to other locations. e Sufficient space to perform repair, maintenance, and service of equipment, displays, and i instrumentation. e Space for ready access to communications equipment by all EOF personnel who need communications capabilities to perform their functions.

Space for ready access to functional displays of EOF data.

e Space for storage of plant records and historical data or space for means to readily acquire and display those records. e Separate office space to accommodate at least five NRC personnel during periods that the EOF is activated for emergencies. 4

                -G ,

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Approved Design Material- Operations page 13.3 7

Synt m 80+ Desian controlDocument 1 i j 13.3.3.2.5 Radiological Monitoring To ensure adequate radiological protection of EOF personnel, radiation monitoring shall be provided in the EOF. These systems shall continuously indicate radiation dose rates and airborne radioactivity ! concentrations inside the EOF while it is in use during an emergency. These monitoring systems shall j include local alarms with trip levels set to provide early warning to EOF personnel of adverse conditions i that may affect the habitability of the EOF. Detectors to distinguish the presence or absence of a radiciodines at concentrations as low as 10 4 microcuries/cc shall be provided. 13.3.3.2.6 Communications l , The EOF shall have reliable voice communications facilities to the TSC, the control room, NRC, and l State and local emergency operations centers. The normal communication path between the EOF and the 1 control room shall be through the TSC. The primary functions of the EOF voice communications

facilities are

.          e          EOF management communications with the designated senior licensee manager in charge of the TSC.

e Communications to manage licensee emergency response resources, e Communications to coordinate radiological monitoring. e Communications to coordinate offsite emergency response activities. { e Communications to disseminate information and recommended protective actions to responsible

government agencies.

f The EOF voice communications facilities shall include reliable primary and backup means of 7 communication Voice communications shall include private telephones, commercial telephones, radio networks, and intercommunications systems as appropriate to accomplish the EOF functions during i @Q emergency conditions.dThe licensee shall provide a means for EOF telephone access to commercial telephone common-carrier services that bypasses any local telephone switching facilities that may be susceptible to loss of power during emergencies. The licensee shall ensure that spare conunercial telephone lines to the plant are available for use by the EOF during emergencies.] 1-j The EOF voice communications equipment shall include: i j e Hotline telephone (located in the NRC office space) on the NRC Emergency Notification System (ENS) to the NRC Operations Center. e Dedicated telephone (located in the NRC office space) on the NRC Health Physics Network (HPN). J

;         e         Dedicated telephones for management communications with direct access to the TSC and the control room.

e Dial telephones reserved for EOF use to provide access to onsite and offsite locations. 3 r iv. S / ) Approved Design Material- Operations Page 13.3-8

System 80 + Design ControlDocument 13.3.3.4 Laboratory Facilities 13.3.3.4.1 Summary Description lf The Laboratory Facilities are out of scope items which shall be provided by the license applica [ Consistent with the requirements stated in NUREG-0654, II.H.9 and NUREG-0737, II.B.3, the System 80+ Standard Plant design makes provisions for inclusion of both hot and conventional Laboratory Q Facilities. Hot facilities are currently located in the Nuclear Annex; additional hot facilities should be I located in a Radwaste Building. Space for a large conventional laboratory should be provided in the 5tj Station Services Building throughout the plant. e tions for other, smaller lab facilities are allocated at various acilities themselves shall be designed according to particular owner 7 pla preference, but they s I s pport efforts to monitor plant systems and environmental samples for compliance with technical s ifications. Complete descriptions of the facilities shall be provided by the licensee. The following sections contain interface requirements which must be met to ensure adequacy with the system 80+ Standard Design. The word "shall" is used to distinguish interface requirements that are mandatory from text that is purely descriptive. 13.3.3.4.2 Function The laboratory facilities shall provide for the following primary functions:

To provide plant support services for routine analyses required for personnel protection, surveys, and related health physics functions, e to provide normal and post-accident cold chemical analyses on required plant chemistry samples, e to provide routine and post-accident counting on all plant radioactivity samples,
to provide grab sample analyses used as a check on the accuracy of the continuous on-line process monitoring instrumentation, and
to provide a facility to store and secure radioactive calibration and check sources and instruments undergoing calibration, maintenance, or repair.

13.3.3.4.3 Location The hot laboratory facilities are currently shown to be located in the Nuclear Annex, and should be included in the Radwaste Building design. Space for a large conventional laboratory should be provided in the station services building. Radiation counting rooms and instrument calibration areas are located at elevation 115+6 in the Nuclear Annex Outage / Maintenance Area. Locations for other, smaller lab facilities are provided at various places throughout the plant. Locations for these facilities are provided l to assure that all critical onsite sampling capabilities (see Regulatory Guide 1.97) can be performed to the required accuracy at the plant site. and such that ALWR normal and post-accident sampling requirements are met. A~ Ft N t

       }           \' . h. Cv Approved Design Material- Operations                                                                  Page 13.3-12

l i i Syntem 80+ Design ControlDocument i i 1 j 13.3.3.4.4 Features i l In order to meet the intent of the ALWR Requirements Document and the aforementioned government l l regulations, the laboratory facilities shall be designed with the following features: ( j

adequate space for expansion to accommodate changes in available technology and equipment, ,

i !

radiation counting rooms, instrument calibration areas and checkout areas located in low radiation f

) zones and provided with shielding to reduce background radiation " noise", and l I : 1

secured access to radioactive calibration and check sources. !

I i' i Liquid wastes from laboratory drains shall be processed by the Liquid Waste Management Systems 'I j (LWMS) described in Section 11.2. Types of waste shall be segregated at the point of origin and routed j to the appropriate LWMS subsystem, e.g., chemical drains accept non-detergent cleaning solutions, ] chemical sample solutions, etc., whereas liquid waste from reactor grade lab drains would be routed to j the Equipment Drain Tank. Due to the cost and complexity of waste disposal considerations, the , I generation of mixed waste shall be avoided whenever possible. 1 j Sampling methods and instrumentation shall be described in plant operations manuals provided by the ] licensee. General maintenance shall be described in other licensee plant operating documents. 13.3.3.5 Post Accident Sampling i

Consistent with the guidance stated in NUREG-0737, II.B.3, the System 80+ Standard Plant design l l provides for a Post Accident Sampling System. This system is located in the CVCS panel. System i

! functions and design requirements are covered in Section 9.3.2. l ! 13.3.3.6 Onsite Decontamination Facilities 13.3.3.6.1 Summary Description f-j [The Onsite Decontamination Facilities (ODF), located in the Nuclear Annex (el. 91+9) and the { Radwaste Facility, are out of scope items which shall be provided by the license applicantMese ; i facilities shall be provided to remove or reduce radioactive contaminants from plant equipment, protective clothing, and personnel. These facilities shall be designed according to particular owner preference, but 4 shall be supplied by the major decontamination equipment, including various spray nozzle assemblies, j chemical and/or abrasive supply systems, collection and s j demineralizers and piping connections to waste processors]grage tan Complete descriptions of the facilities shall be provided by the licensee. 1 The following sections contain a description of a typical ODF and interface requirements which must be j i met to ensure adequacy with the System 80+ Standard Design. The word "shall" is used to distinguish ! j interface requirements that are mandatory from text that is purely descriptive. i l ! The ODF shall include the hot laundry facilities, hose washdown stations, personnel decontamination j fixtures, hot shower, radiation detection equipment and personnel decontamination supplies. Also j equipment necessary to decontaminate small tools and instruments as well as larger tools and pieces of

equipment shall be provided.

l A

         $      CJL ;a[o ,-

. Agnproved Design Meterial . Operations Pope 13.3-13 h bM b ***

N*M N + Deskn ControlDocument 13.3.3.6.5 Decontamination Methods and Procedures (( Selection of decontamination methods to be employed in the Onsite Decontamination F specific generating plant is the responsibility of the individual licensee Some of the decontamination requirements may be met by using portable or otherwise transportable facilities at the discretion of the individual the licensee. licensee. Description of those methods and rationale for their selection shall be p Decontamination and radwaste control procedures are considered to be a fundamental part of the operations doc mentation. The licensee plant operations documents shall contain these detailed procedures.] All decontamination methods and procedures should be consistent with accepted industry practice, and shall meet the intent of the referenced section of the Code of Federal Regulations. l 1 l l { (DV th.~ Apprend Design Meterial Operations !*9* 13 3*I5

Syst:m 80+ Design ControlDocument 13.3.3.6.5 Decontamination Methods and Procedura (( Selection of decontamination methods to be employed in the Onsite Decontamination Facilities at a specific generating plant is the responsibility of the individual licensee Some of the decontamination requirements may be met by using portable or otherwise transportabh racilities at the discretion of the individual licensee. Description of those methods and rationale for their selection shall be provided by the licensee. Decontamination and radwaste control procedures are considered to be a fundamental part of the plant operations doc mentation. The licensee plant operations documents shall contah these detailed procedures.] All decontamination methods and procedures should be consistent with accepted industry practice, and shall meet the intent of the referenced section of the Code of Federal Regulations. f CDL (.h. l Approved Design Material- Operations Page 13.3-15

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System 80+ Deskn Crntrol Document 13.4 Review and Audit (Dnformation concerning th.3 site operator's reviews and audits is within the site operator's scope and shall be provided in the site-specific SAR. f- , r h h Ld , [ w eo+u. .e.n !*oe 13.4-t

Swtem 80+ Desfort ControlDocument 13.5. Plant Procedures I hinformation concerning the site operator's plant procedures is within the site operator's scope and shall o be provided in the site-specific SAR.)) 1 ; 13.5.1 Plant Operating Procedmus Development Plan ' i e A Plant Operating Procedures (POP) development plan '(i.e.', the Plan) shall be created to l formally guide development of the POPS for normal, abnormal, and emergency operations. ! e The Plan shall specify the process by which the POPS will be developed, verified, validated,

                                                           '                                                         f' revised, and maintained.

e The scope of the plan shall include the applicable operating procedures in Subsection 13.5.1.1 e The Plan shall ensure that POPS follow standard formats implemented through Writer's Guides (e.g., per References 1 and 8). e The Plan shall specify that the POPS' technical bases are documented to ensure that their content l is consistent with plant design basis material including applicable procedure guidelines, task analyses, and PRA, (all provided via OSIP), and applicable plant-specific aspects, per Reference 1. e The Plan shall specify the process by which training of operators on the POas and on changes to the POPS is provided and kept current. t i e The plan shall specify the POP validation ac 'vities that demonstrate the acceptability of the - ; completed procedures for the scope specified in item C)bove. e

                                                                           'R The plan shall specify that the POPS comply with the applicable requirements of References 1 through 7.

l 13.5.1.1 Procedures included in Scope of Plan l l The following procedures shall be included in the scope of the Plant Operating Procedures Development i Plan: System Procedures Such procedures include all system procedures that require operator action in the MCR or RSR. l Procedures shall be prepared for PWR system operation. Typical PWR systems having procedures are I listed below: e Reactor Coolant System l e Control Rod Drive System (including part-length rods) ' e Shutdown Cooling System I e Emergency Core Cooling System i % COL In h . . . l 4pwont Deafpn nieteniel- Opereelant Pnee 13.5-1 l i

SENT BY:DEAS t11-23-34 ; S:30 : DUKE EMR & SMyd* .* o System 80+ oneaen coneer Docennent e Hot Standby to Cold Shutdown e Reactor coolant system operation with loops partially drained n= diad- Control Prer=dures Such procedures include all radiation control procedures that require operator action in the MCR or RSR. 7)ical radiation control procedures are listed below: V e PWR Gaseous Effluent System Y 1 Collection, Storage, and Discharge

2. Sampling and Monitoring

3. Air Ejector and Stack Monitoring 4 Ventilation Air Monitoring I

e Process radiation Monitoring System Operation e Meteorclogical Monitoring Mainan===nce, Calibration Inspectica and Test Proerdures All portions of maintenance, calibration, la=~*ina and test procedures that require operator action in the MCR or RSR will conform to the requirements of the Plant Operating Procedures Development Plan. Prarwhnu for Esmergemeiss, Operational Transiasts and Other Signifiennt Events Such procalures include emergency, operational transient and other significant-event-related procedures that require operator action in the MCT or RSR. Typical procedures for emergencies, operational transients and other significant events are listed below: e Loss of Coolant (including significant PWR steam senerator leaks), (inside and outside primary I containment), (response to large and small breaks, including leak rate determination) e Loss ofInstrument Air e Ims of Electrical Power (or degraded power sources, or both) e Loss of Core Coolant Flow / Achievement and Mairweamare of Natural Circulation, including connection of the pressurizer heaters to the emergency bus, if =- es y e Loss of Condenser Vacumn e Loss of Service Water

IAss of Shutdown Cooling
Ims of Component Cooling System and Cooling to Individual Components

         ' a s g.as.am w -cys esas vs.s4

l i Svotem 80+ Design ControlDocument

8. Techniques for Preparing Flowchart Format Emergency Operating Procedures. NUREG/CR-5228, 1989.

13.5.2 Administrative Control Procedures ((Such procedures will be a COL action item including " Review and Modify Procedures for Removing Safety-Related Systems from Service" TMI Action Plan Item II.K.1(10), " Guidelines for Upgrading Other Procedures," and " Revise Deficiency Reporting Requirements," TMI Action Plan Item II.L4.1.g To avoid excessive fouling and plugging of the screens near the IRWST suction inlets during an accident (see Section 6.8 and Appendix 19.8A, Subsection 2.9), the containment must be cleaned of sand, l maintenance debris, and other particulate materials prior to startup from a refueling outage. There are no specific requirements for determining acceptable cleanliness, but maintenance practices should be consistent with generally accepted practices for nuclear power plant containment interiors. This is a DRAP item; see Section 17.3. l The COL applicant / holder shall consider the guidance of NUMARC-8700, " Guidelines and Technical . Bases for NUMARC Initiatives Addressing Station Blackout at Light Water Reactors," (11/20/87) in { developing plant procedures for severe weather conditions such as hurricanes and tornadoes. l f O h 6l fu Approved Design Materlaf- Operations p ,,,y3,5.g

Syotem 80+ oesign controlDocument 13.6 Industrial Security h Information concerning the site operator's industrial security is wjthin the COL applicant's scope and will be presented by the COL applicant in a separate document.jp 13.6.1 Interface Requirements Site security is an out of scope item which shall be provided by the COL applicant. The COL applicant shall verify that the following interface requirements are met to ensure adequacy with the System and Standard Design: e The design of the physical security system shall be in conformance with 10 CFR Part 73, 73.55 including:

1. 10 CFR Part 73,73.55 in its entirety and Part 73 Appendices B and C

2. 10 CFR Part 25 and 10 CFR Part 95

3. Regulatory Guide 5.44

4. NUREG-0674

5. 10 CFR 73.56

6. 10 CFR Part 50, 50.70 (b) (3)

7. Regulatory Guide 5.12

8. Regulatory Guide 5.20

9. Regulatory Guide 5.65

10. Regulatory Guide 5.66 e A comprehensive listing and evaluation of all vital equipment, vital piping, vital power sources, vital water storage facilities, etc., shall be developed (A star.dard listing of vital equipment within the standard Design scope is given in Appendix 13B). This Listing and evaluation should include all support functions vital to equipment operation (e.g. diesel generator cooling water, D/G fuel, HVAC considerations for vital electrical equipment, cooling water for component cooling, etc).

This listing and evaluation shall be controlled in accordance with current security safeguards information (SGI) procedures and guidelines. e The design of the security system shall include an evaluation of its impact on plant operation, testing, and maintenance. This evaluation shall assure that the security restrictions for access to equipment and plant regions are compatible with required operator actions during all operating and emergency modes of operation (i.e., loss of off-site power, access for fire protection, health physics, maintenance, testing and local operator), and shall include an evaluation of access control system failures. j jy i Ak3.* Approved Design Material Operations Page 13.5-1

     ~. --- .                        --.               .     - = .           = .--   . _ . , .           . . _ - ._         - _ _ _ _ _ _ _ _ _

Syntem 80+ Deslan ContwlDocument ! physics, maintenance, testing and local operator), and shall include an evaluation of access control system failures.

Securit'; system facilities and equipment (e.g, PAP, CAP, security computer (s), security communications (fixed and portable) equipment, security power system, etc.) shall be located within the protected area to assure that the security force can respond to security events.

J

The security communications subsystem shall meet the following requirements:

1. Each on-site security officer, watchman, or armed response individual shall be provided 1 with continuous communications with an individual in each continuously manned alarm )

station (i.e., CAS SAS, PAP). This may be accomplished by using multi-frequency radio or microwave transmitted two-way voice communications.

   ,                          2.         Communications shall be provided between the main plant control room and nuclear plant CAS and SAS (i.e., dedicated telephone service that does not have any terminations                      !

outside the protected area boundary, radio, etc.).

b sabotage vulnedoility analysis shall be performed by the COL Applicant on the final site-specific design khich shall include:

1

1. A review of the sabotage vulnerability of vital safety components to tampering by an insider with authorized access.

l 1 l 2. An evaluation of whether timely means exist to discover and compensate for such tampering. J

3. The incorporation of design or procedure changes, if practical, which could eliminate l opportunities for tampering or to ensure the ability to compensate for tampering.]i-t Appendix 13A shall be utilized as the starting point for the sabotage vulnerability analysis.

4 i Co( - AMworedDesko Atatoriel Operations Pege 13.6-2 _ . - . - , _ . ~ _ . . - . _ ,_

system 80+ oesign contros oocument 8.0 Evaluation of Damage Control Measures An evaluation of the damage control measures (DCMs) provided in the System 80+ Standard Design was conducted. The approach taken followed that in Reference 2. Reference 2 identifies 18 DCMs for pressurized water reactors. The 18 DCMs were reviewed against the System 80+ Standard Design. 8.1 Summary of Damage Control Measures The damage control measures evaluated are identified below (and numbered consistent with the Reference 2 listing): Damage Control Meas INumber Description of DCM DCM 3 Modify the main feedwater system turbine pump train to operate following a loss of offsite power and use as a substitute for the EFW system. It is assumed the EFW has been successfully sabotaged. DCM 4 Provide a cross-connect from Safety Injection to substitute for EFW system. It is assumed the EFW has been successfully sabotaged, and a loss of offsite power has made the main feedwater system unavailable. DCM5 Provide an redundant set of main stearu overpressure system safety relief valves to vent steam for SG depressurization and heat removal. It is assumed that the main steam overpressure protection (ADV's and safety valves) has been successfully sabotaged. DCM 9 Modify the Safety Injection System to use tandem SI pump for injecting water into the RCS at " full" RCS pressure. It is assumed successful sabotage prevents the depressurization of the RCS to injection pressure. (NUREG/CR-2585 considers this DCM feasible but characterizes implementation as questionable.) DCM 13 Modify plant (non-safety) service water system to operate following a loss of offsite power and cross-connect to substitute for Essential Service Water (ESW) system. It is assumed the ECW pumps have all been successfully sabotaged. DCM 14 Modify condensate system to operate following a loss of offsite power and cross-connect to substitute for ESW system. It is assumed the ESW pumps have all been successfully sabotaged. DCM 15 Modify the Fire Water System to be fully operable following a loss of offsite power and cross-connect to substitute for ESW system. It is assumed the ESW pumps have all been successfully sabotaged. DCM 16 Provide a cross-connect for ESW system to provide direct cooling of components serviced by the Component Cooling Water (CCW) system. Approved Design Material- Operations Page 13A-14

Syst m 80+ Design ControlDocument Damage ConM MeasuSNumber Description of DCM DCM 17 Provide local indicators for pressurizer and steam generator level to be available upon loss of offsite power and successful sabotage of the Process Instrument I Cabinets. Not all of the transmitters are assumed to be affected by this sabotage. Only remote indications are affected. DCM 18 Provide local indication of steam generator pressure to be available upon loss of all Process Instrument Cabinets. Not all of the transmitters are assumed to be affected by this sabotage. Only remote indications are affected. DCM 19 Provide the ability to reenergize Non-Class IE loads from a Class IE AC system. It is assumed that the IE systems (e.g. EFW, ESW, etc.) to maintain the plant have been successfully sabotaged. 1 DCM 20 Provide to the DC electric supply system the ability to have bus tie capability l between Class IE DC load groups. It is assumed that part of the DC electric supply system has been successfully sabotaged. DCM 21 Provide the ability to reenergize Class-lE DC loads from Non-Class IE DC load I groups. It is assumed that the Class IE DC supplies have been successfully sabotaged. DCM 22 Provide the DC electric power system with a load transfer capability for selected DC loads. It is assumed that the DC supply has been successfully sabotaged. DCM 23 Modify for rapid alignment of backup water sources for the purposes of decay heat removal or coolant inventory control in the RCS. It is assumed that the alignments for normal inventory sources have been successfully sabotaged. DCM 25 Modify the EFW system for operation without AC or DC power. It is assumed the electric supplies for the EFW components have been successfully sabotaged. DCM 26 Add an alternate AC electric source onsite for Non-Class 1E power. It is i a.csumed that various Non-Class IE components are desired as the result of ! successful sabotage to Class IE power systems or safety-related systems required i for maintaining the plant in a safe condition (e.g. EFW, ESW, CCW, etc.). I DCM 27 Modify the diesel generator for startup and loading without an AC or DC power source. It is assumed that successful sabotage has made AC and DC power sources for diesel startup inoperable. 1 1 Approved Desigrr Meteriel- operettorus Page 13A.15

l l l . l Syst m 80 + Design ControlDocument l l l l 8.2 Summary of Sabotage Scenarios l l The sabotage scenarios used to evaluate DCMs are identified below: ;

i

General Scenarios Characteristics

1. A majority of the scenarios assume a loss of offsite power coincidentally with the successful sabotage of a safety system. This event results in a main turbine trip due to a loss of load. Therefore, components powered by the Non-Class IE AC power are unavailable. Unless affected specifically by the sabotage the emergency diesel gener tors !

operate and supply AC power to the Class IE buses. l l By the nature of the general event it is assumed Reactor Coolant Pumps, main feedwater and condensate systems, normal pressurizer control, and steam generator control ! functions are not available. f e Specific Scenar o' Sabotage Evaluation [of CESSAR Dd This review determined that all l l PWR DCMs ret i

1. Emergency Feedwater (EFW) system components are assumed to be successfully sabotaged and unavailable for steam generator inventory control. [DCM 3,4,9,19] & Wg g l y og !

2. Essential Service Water (ESW) components are assumed to be successfully sabotaged I making the systems it supports unavailable in a short period of time. Specifically Component Cooling Water (CCW) and the components it serves. [DCh # 13,14,15]

3. Main steam overpressure protection valves (ADVs and safety valves) are assumed to be (e successfully sabotaged and unavailable for steam generator depressurization. [DCA -

5] !

4. Component cooling water pumps are assumed to be successfully sabotaged making them unavailable to satisfy their design function. [DCh 16]

5. V Remote indication for key parameters affecting heat removal (specifically pressurizer level, steam generator level and pressure) are assumed to be successfully sabotaged so that they can not be available to assess plant tatus from remote locations (e.g. the control room or remote shutdown panel). [DCM # 17, 18] Y

6. Class IE AC power is assumed to be successfully sabotaged such that a minimum of one train of safety-related equipment is not available. [DCM 19,26,27]

Y

7. Class IE DC power is assumed to be successfully sabotaged such that a minimum of one train of IE AC and DC safety-related equipment are not available. [DChh:

20,21,22,27] % b

8. , pid alignment of inventory sources are assumed to be successfully sabotaged such that available safety related components do not have the condensate inventory to meet the intended design functions. [DCM]: 23]

D ( Approved Design Material- Operations Page 13A 16

Syst m 80+ De-Ign ControlDocument 8.3 Evaluation of Damage Control Measures and Scenarios The Damage Control Measures (DCMs) listed in Section 8.1 above were evaluated for the System 80+ Standard Plant design taking into consideration the PRA insights and assumptions in Section 19.15,of : SAR:DC' This review determined that all PWR DCMs are applicable to the System 80+ design. l f The DCMs were then evaluated against the scenarios identified in Section 8.2 above and the transient sequences analyzed in Sectiors19.4 and 19.9 of the System 80+ PRA. j A The evaluation of the DCMs against scenarios and accident sequences revealed that the System 80+ design incorporates the necessary DCMs required to mitigate the scenarios. Table 13A-1 provides a summary of the evaluation of each DCM in Section 8.1 above. , ,. The System 80+ PRA results and insights are summarized in Section The PRA 19.15,bf C demonstrates that System 80+ has achieved a low core damage frequency and low risk through a balance of prevention and mitigation design features. The System 80+ design uses redundancy and diversity to assure maintenance of the safety functions for prevention and mitigation of transients and accidents. Quadrantization with physical separation and barriers between the equipment in the different trains and divisions also provide protection against physical challenges such as fire and flood. The design features which make System 80+ a safe, reliable, and robust plant also provide substar.tial protection against inside sabotage. 9.0 Conclusion Front-line safety systems are provided with sufficient indications and alarms and are physically separated l I such that complete undetected disabling of a safety system is highly unlikely. Indications are provided in the control room to alert operators of system status. Position indicators and alarms are provided on safety system manual isolation valves (See Section 6.0) and remotely operated valves to detect mispositioning. Indicators and alarms are provided in the control room to indicate safety system / component inoperable / bypassed conditions. Flood doors and doors to vital I&C equipment rooms and cabinets have annunciators or indicators to detect unauthorized entry to safety system equipment areas. Safety system and associated equipment and components are physically separated by division and channel. Multiple rooms and cabinets must be entered to disable a safety system or function. Together, all of these features provide ample indication of tampering and time to respond and foil an insider's threat to completely disable the plant's safety systems. 1 Adequate design features are provided to prevent an insider from causing a LOCA. The SCS suction line is afforded protection from sabotage in that the isolation valves are located inside containment, requiring the saboteur to gain control of the valves from locations outside containment to cause a 'areach in the suction line. Valve position indicators and alarms and spatial separation of these electric controls provide detection and ample time to respond to the threat. Additionally, the SCS system is designed with a low pressure nominal rating of 900 psia. Likewise, the SDS and RCS safety valves are located inside containment with controls outside containment. Multiple rooms and cabinets must be entered to successfully gain control or disable these valves. The RCGVS valves are afforded the same protection features as the valves described above. Additional instrumentation is provided downstream of the RCGVS valves to detect RCS leakage which will alert the operator of an open valve. The CVCS let lown line outside containment is provided with sufficient control room indications to detect a breach in the line. Approved Design Material- Operations Page 13A-17

im m 2 u%m. m -

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I System 80+ Design ControlDocument h 14.0 Initial Test Program ak 14.1 Specific Information to be Included in PSAR g-- __ This section is not applicable to tw c'f y An'1yde "grt.' See Section 14.2 for a description of the initial test program. ./ Oc 6 f.m ' Pao & Ag0rovd ledin MakC M .

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i v;p CESSAR EHL"icuc. 2 me -~ -a %

                               % ~ y- As a Gu cxc7 1, 14.2.12.3.2         Isothermal Temperature Coefficient Test             i 1.O        OBJECTIVJ 1.1       To measure the Isothermal Temperature Coefficients (ITCs) for various RCS temperatures, pressures, and CEA configurations."

l 1.2 To determine the Moderator Temperature Coefficient (MTC) from the measured ITC. PREREOUISITES 2.0 2.1 The reactor is critical with a stable boron concentration and the desired CEA configuration and RCS temperature and pressure. 2.2 The reactivity computer is operable. 3.0 TEST METHOD 3.1 Changes in RCS temperature are introduced and the resultant changes in reactivity measured. 3.2 Reactivity and power swings are limited by compensation with CEA motion when necessary. 4.0 DATA REQUIRED 4.1 Conditions of the measurement: 4.1.1 Pressurizer pressure 4.1.2 CEA configuration 4.1.3 Boron concentration 4.2 Time dependent information: 4.2.1 Reactivity 4.2.2 CEA position 4.2.3 Temperature uw-For unit, this test will be performed only at-56587,

            -2250-psia. oca err.oc    rc e swrse4 wo hmM -

Amendment E 14.2-267 December 30, 1988

1 System 80+ Design Control Document

                                                                                                  \-   4        . - -

Table 14.3-2 Probability Risk Assessment Paragraph Assumption / Parameter Description /Value "I 19.6.3.1 The main power system consists of the main generator, associated isolated phase buses, generator circuit breakers, three single phase unit main transformers, two unit auxiliary transformers, and two reserve auxiliary transformers.

                                         "I 19.6.3.1                        The 4.16 Kv non-Class IE power system consists of four switchgears                            ;

and a non-Class IE alternate AC source. Two of the switchgears (i.e, l X - non-safety switchgear and the X rm'ari$nt non-safety i switchgear) are powered by one of the unit auxiliary transformers, l while the other two switchgears (i.e., Y - non-safety switchgear and the l Y - permanent non-safety switchgear) are powered by the other unit auxiliary transformer. The permanent non-safety switchgears can also be powered by the reserve auxiliary transformers and the alternate AC source.

                                        "I 19.6.3.1                         Each division of the 4.16 Kv Class IE power has two switchgear.

Both switchgears within a division are typically powered from their associated 4.16 Kv permanent non-safety IE switchgear. Both ; switchgears can also be powered by the emergency diesel generator of the same division.

                                                                                                                                            )
                                        "I 19.6.3.1                        Each division of the 120 VAC C s IE power system has three inverters and associated buses. Each inverter within the division is                s powered from a separate 125 VDC Class IE bus in the same division of the EDS.

y ' l A 19.6.3.1 "I Each division of the 125 VDC Class IE power system has three battery , Dv , 3, L chargers, three batteries, and three distribution centers (buses). Each battery charger is powered from a separate 480 VAC Class IE motor Y yJ 2D control center within the same division of the EDS. Each battery is sized to supply its emergency loads for a minimum of 2 hrs without fyd c recharging. y (,}j "I 19.6.3.1 The emergency diesel generators are physically and electrically isolated from each other. ih N / , c I pl V 19.6.3.1 The starting air storage capacity for each emergency diesel generator is sufficient for starting the diesel generator for a minimum of five times. g )% s I "I 19.6.3.1 Each emergency diesel generator has a complete and separate fuel oil storage system. The storage system has sufficient fuel that allows the

                                                                                                                             .g           f emergency diesel generator to operate supplying post DBA-LOCA                   h6w 9y7 loads for a time period of no less than seven days.                                      -O 19.6.3.1                   "I                                                                                             S3 Each emergency diesel generator is automatically staned and loaded by                    '1 t

the Engineered Safety Feature - Component Control System %y ~%y C "I (ESF-CCS).

                                                                                                                           \Y         3 h

19.6.3.1 Each ernergency diesel generator is housed in Seismic Category 1 / structure to guard against earthquakes, fires, and missiles. Approved Design Material- Initial Test Program page 14.319

 - w- = -                     -
                                                                                                                                 ~

System 80+ Deelan coneret Docenment 15.0.3.3.3 Shntdown CEA Reactivity The shutdown reactivity is dapandaar on the CEA worth available on reactor trip and the axial power distribution. For most transient analyses, conservative total CEA worths of 8.86 percent and 6.5 percent As were used for hot full power (HFP) and hot zero power (HZP), respectively. For some events, more conservative values were used (i.e., less negative). However, in the case of steam line break events a CEA worth of 10% was used for the full power cases. This worth is consistent with the moderator reactivity versus nuderator temperature function used in these analyses (see Section 15.1.5). The foregoing values include uncertamties, the most reactive CEA stuck in the fully withdrawn position, and the effect of % on CEA worth for events initiated from HZP (Section 4. 2.4.3). The shutdawn reactivity worth versus position curve which is employed in the majt. ity of Chapter 15 analyses is shown in Figure 15.01 and is applicable for an axial shape with an Axial SL pe Index (ASI) of +0.3. 'Ihis shutdown worth versus position curve yields a conservatively slower rate of negative reactivity insertion than is aWd to occur during the majority of operations, inchiding power ) maneuvering. Accordingly, it is a conservative representation of shutdown reactivity insertion rates for reactor trips which occur as a result of the events analyzed. For some events, a dynamic axial power function is utilized based on the HERMITE Code (see Section 15.0.3.1.7). 15.0.3.3.4 EKective Delayed Neutma Fraction l The effective neutron lifetime and delayed neutron fraction are functions of fuel burnup. For each

analysis, the values of the neutron lifetime and the delayed neutron fracuon are selected consistent with i the time in life analyzed.

15.0.3.3.5 Decay Heat demeration Rate l Analyses assume decay heat generation based upon infinite reactor operation at the initial core power level identified for each event.

15.0.4 Radiological Consequences The objective of the radiological evaluations is to confirm that the calculated doses from postulated accidents lie within the limits described in 10 CFR 100.11 and/or the applicable (5, NUREG 0800 sections. Doses are dependent, in part, upon the assumed values of radiological 'g l atmospheric dilution, x/Q (see Section 2.3). Values assumed for Chapter 15 analyses were

! determined using meteorological data representative for 80-90% of existing US reactor sites. Several of the non-LOCA events di-=ad in Chapter 15 are accompanied by the release of l steam or liquid from the reactor coolant system or main steam system. The CESEC computer code (described in Section 15.0.3.1.3), in combination with hand calculations, were used to j i determine the mass and energy releases as a function of time. v i m - .p . , , _ l The Combined License Applicant for each site will calculate site specific x/Q using the

l meteorology, exclusion area boundary and low population zone for that particular site to show that the x/Qs assumed in the Chapter 15 analyses are boundingg l l h (,. [ A tyWe 6\ vw , 4 te 1eb T/thc cL,cbh Sedw 3.2 . m Approved Desen AfeestdsI- Acaddset Ano$ses Aspe fs.0-9

gr ~ - e= Design Convol Document System 80+

51. "The ROCS and DIT compmer Codes for'N0 clear Design", CENPD-266-P-A, C-E Proprietary TopicalRepon, April 1983.

52. I2ner to A. E. Scherer (ABB-CE) from Cecil O. 'Ibomas (NRC)", AEW-am for Rr.fum iug of Iirming Topical Report CENPD-266-P CENPD-266-NP "The ROCS and DIT Cewnr=w-Codes for Nuclear Design" April 4,1983. l

53. Report: CEN-133(B), " FIESTA: A One Dimensional, Two Group Space-Tune Kinance Code for Nenlaring PWR Scram Reacavities", November 1979. Approval: letter, R. A. Clark (NRC) to A. E. Lundvall, Jr. (BG&E) Docket Nos. 50-317 and 50-318, Approval of CEN-123(B), March 13,1981.

54. "C-E Methods for Loss of Flow Analysis", CENPD-183-A, '

June 1984. (NRC .pyivval leuer

~~

dated May 12, 1982).

55. Draft NUREG-1465, "Arndent Source Terms for Light-Water C Nuclear Power Plams3 auno992;>

56. " Licensing Design Basis Source Term Update for the Evolutionary Adviaal Light Water Reactor", Advanced Reactor Severe Aendent Program (ARSAP) Source Term Expert Group, S 990.

57. NUREG4800, "U.S. Nuclear Rent =mry Commission Standard Review Plan, Section 6.5.2, Cw-" - = Spray as a Fission Product Cleanup System". Dvder 1988 (Revision 2).

58. NUREG-0800, "U.S. Nuclear Regulatory Ci- don Standarti Review Plan, Secnon 6.4, Control Room Habitability Systems",1981 (Revision 2).

59. F. J. Moody, " Man ="n Flow Rate of a Single-Cuioyesem, Two-Phase Mixture," Journal of Heat.Transfgr. Trnnaerions of the ASME, F 1%5.

 "Y         ,                                                _

GC.* NUREG/CR-4691, Vol. 1; "MELCOR Accident Consequence Code M/ ' " Guide"; Sandia National G y System (MACCS) Laboratories; February 1990. User's Gl. NUREG/CR-4691, Vol. 2; "MELCOR Accident Consequence- Code System (MACCS) - Model Description"; Sandia National y Laboratories; February 1990. Q. NUREG/CR-4691, Vol. 3; "MELCOR Accident Consequence Code V Programmer's Reference

Manual"; Sandia System (MACCS) -

National Laboratories; February 1990. j (o "6, NUREG/CR-4551, Vol. 2, Rev.1, Part 7; " Evaluation of Severe V Accident Risks: Quantification of Major Input Parameters - s MACCS Input", Sandia National Laboratories; Decemberas - 3 1990. Pope !$.0 14 Anraveer Design Atosenet. Acodont Ana4see

c M - System 80+ Deakn coneraloccumerrr

                                        *Ihe opemng of an ADV increases the rate of heat removal by the steam generators causing cooldown of the RCS. Due to the negative moderator reactivity coefficient, core power increases from 102% of rated core power, reaching a new, stabilized value of 115%. The feedwater control system, which is assumed to be in the automatic mode supplies feedwater to the steam generators such that the steam generator water levels are ==i==iaad-HM (I tco ,2.R)

Dunng the IOSGAD transient, a R of 1.36 occurs just before the operator manually trips the or at 1 *At 1800.55 seconds the trip breakers open, ne ws h i. turbine trip is ini+=*ad loss of offsite power is ====nad to occur and the

      ,>                            -   to__ coast down.1At QgM}aarewwin the steam generator water level reaches the gA7                               emergency feedwater acmation setpoint and -awy feedwater delivery begas wnhin 60 seconds.fDue to the RCPs enaardawn, the transient DNBR d w,                                i sQ a animmum value of 1.30 at 1801.96 seconds and then rapuity increases as                         j shown in Figure 15.1.4-1.15. From 1805'macanda to 1845 narande the MSSVs release steam. At 1919.46 seconds a void begms to form in the RV upper head.

At 2050.4 seconds the steam generator pressure drops below the MSIS setpoint of 719 psia. The MFIVs and MSIVs close by 2056.75 seconds. Safety injecnon is iniciarad at 2115.5 seconds. 'Ihe affected steam c 4-i dries out at 2154.6 i seconds. At 3000 seconds the operator manually closes the open ADV. The I operator initiates plant cooldown at 3600 seconds. I Case 2: Inadvertent Openmg of a Steam Generator Atmospheric Dump Valve with Imss of the Feedwater Control System Reactor Trip Override (IOSGADV + SF) and with i IAss of Offsite Power. The dynamic behavior of the salient NSSS parameters following IOSGADV with loss of offsite power and with loss of the Feedwater Control System Reactor Trip Ovemde is presented in Figures 15.1.4-2.1 through 15.1.4-2.15. Table 15.1.4-2 summarizes the major events, times and results for this transient. The opemng of an ADV increases the rate of heat removal by the steam generators causing cooldown of the RCS. Due to the negative moderator reactivity coefficient core power increases from 102% of rated core power, reaching a new, stabilized value of 115%. '1he feedwater control system, winch is assumed to be in the automatic mode, supplies feedwater to the steam generators such that the steam generator water levels are ==i==inad. Durmg the IOSGADV + SF transiem the operator manually trips the reactor at 1800 seconds, at which time the DNBR is 1.36. At 1800.55 seconds the trip breakers

 '                                        open, turbine trip is initiated, loss of offsite power is assumed to occur and the RCPs -

begm to coast down. At 1800Bseconds the steam generator water level reaches the emergency feedwater acmation setpoint and emergency feedwater delivery begins within 60 seconds. Due to the RCPs coastdown, the transient DNBR decreases, reachinF a tninimum value of 1.29 at 1802.15 seconds and then rapidly increases as i shown in Figure 15.1.4-2.15. The 10SGADV event plus the limiting SF does not j result in fuel pins in DNB. Voids begin to form in the upper head of the reactor vessel at 1829.7 nds. it 1

                                                                                                                         , o. . .. ,
                 ,s                   as.-.         m.-,

i

~-

3,

i I System 80+ Denen Control Document Table 15.1.4-1 Sequence of Events for Full Power Inadvertent Opening of a Steam Generator Atmospheric Dump Valve (IOSGADV) with Loss of Offsite Power Tlase (see) Event Setpoint or Value 0.0 One armaarhenc danp valve opeos fully 1800 Hot channel DNBR 1.36 1800 Operssor initima*4 manual trip 1800.4 Manual mactor trip signal generated

                                                                                                                        -                     b
                  )                                     Ranctor trip breakers opentrurbine Trip / Loss 1410.55 NhN  N J Egin m c A                                                                                                    ,

Steam generator water level reaches emergency 19.9 % , 1800.22 { q 1 1801.96 feedwater armarian analple setpoint. %WR Minin== Tranment DNBR

                                                                                            "        ~

1.30

                                                                                                                                          ~
                                                                                                                                              )

1805 Main steem safety valves open psia 1212 1845 Main steam safety valves close, psia 1151

                                                                                                                         -                             i 1860.22         Emergency feedwater delivered to generator 1919.46        Void begas to form in RV upper head 2050.4         Sseem generator pressare reaches main steam                      719 isolanon signal MSIS) analysis seapoint, psia                                                  )

2056.75 MF1Vs close ea pl*Iy - f 2056.75 MS1Vs close _- ,' 'y - l l 2074.9 Pressuruer pressure reaches low pressurizer ,, 3 1555,, , ! safety injection =emanian analysis setpost. ! Psia 2115.5 Safety injecnon pumps reach full speed - 2154.6 A5ected steam generator dries out - 3000 Operator manually closes ADV 3600 War initiates plant cooldown - W p Ceykr

                     <* A - h o p                                                                                                                      ;

m Angerewed Deste Asseerduf. AseWest Anahese page rf f.2o n .

i @ ) System 80+ Deenan Coneof Documart i l Table 15.1.4-2 Sequence of Events for Full Power Inadvertent Opening of a Steam j Generator Atmospheric Dump Valve with Loss of the Feedwater Control System Reactor Trip Override (IOSGADV+SF) and with 1 j IAss of Offsite Power 4$,tk i

,,V

)! Time (sec) Event M l s 1 ( or Valop j 0.0 One %c dump valve opens fully - 1800 Hot channel DNBR 1.36 1800 Opermor imtiates manual trip Manual reactor trip signal generated - I 1800.4 { j 1800.55 Reactor trip breakers open/ Turbine Trip / 1 Loss of Offsite Power /RCPs begin to l 3- --_Q.. --- ~ 1 1800.6 Steam generator water level reaches 19.9 l emergency feedwater actuation analysis, J setE94!t, %WIL, V 1.29 i ,._,1,802.15 Mhtranswww DNBR l ((829.76 y Void begas to form in RV upiier hemT] - l W ^ ; Q y, , p_ --- ' ' - 719 i steam isolation signal (MSIS) analysis j seapomt, psia l 1848.15 MFIVs close completely - l 1848.15 MSIVs close E- _,'cly - 1860.6 Emergency feedwater delivered to genermor r-

1872.8 Pressunzer pressure reaches low 1555 l pressunzer pressure safety injection j actuanon analysis setpoint, psia i 1913.4 Safety injecuan pumps reach full speed -

1 3000 Operator manually closes ADV - l 3600 ) Operator initiates plant cooldown - 1 l ~

!                                       N I -SKxop                                                                               JM.

i . N 1 ) I i l 4prowssf DenW Asseerdef AceWent W Phee f f. F-2 7 i

Svatem 80+ oesian control Document Table 15.1.5-12 Parameters Used in Evaluating the Radiological Consequences of Steam Line Breaks Outside Containment Opstream of MSIV vaine Parameter SLBFPLOPD (Case 5) SLBZPLOPD (Case 6) A. Data and Assumptions Used to Evaluate the Radioactive Source Term l Power Lael, MWt

1. 3992 10

2. Burnup, MWD /MT 28,000 28,000

3. Percent of Fuel Assumed to 0.5 0 ,

Exponence DNB, % , l

4. Reactor Coolant Activity Before Tech Spec Appendix 15A Tech Spec Appendix 15A Event (based on 3992 MWt),

5. heaadary System Activity Before Tech Spec Appendix 15A Tech Spec Appendix 15A Event l

6. Primary System Liquid Inventory, 638,000 638,000 lba

7. Steam Generasor Inventory, Ibn Affected Steam Generssor 108,640 414,386 Intact Steam Generator ; 108,640 l 414,386 B. Data and Assumpoons Used to Esumase Activity Released from the Secondary System l I

1. Primary to Secondary le.ak Rase, 1.0 (total) 1.0 (total) om _ _

, 2. Total Mass Release from the 390,050 570,520 l Affected Steam Generator, Ibn (0-30 min) l 3. Total Mass Release from the Insact 1,350,990 (2 hre) 1,351,950 (2 hn) ! Steam Generator 2,885,400 (8 hrs) 2.885,440 (8 hn) l l 4. Percent of Core invensory of 5 N/A Volatile Fission Products Assened l % in k Gap

5. Iodme/Casam/ Rubidium 1.0 1.0 rwcane=h Factor in the Affected Steam Generssor I

k @~gd Ob C m 1 k -3 + -" ? A O e2

th - _ _ w J I\m M v s W Except for cases assuming pre-existing and concurrent iodine spike. Amaroved oneinn neenner . Acedent Anetroen Pope 15.132

i l System 80+ Declan Control Document 1 Table 15.1.5-12 Parameters Used in Evaluating the Radiological Consequences of Steam Line Breaks Outside Containment Opstream of MSIV l (Cont'd.) j Value 1 i Parameter , SLBf?LOPD (Case 5) SLBZPLOPD (Case 6)

6. Iodme/ Cesium / Rubidium 100 100

] Decontammation Factor in the j intact Steam Generator 4 7. Peabag Factor 1.58 N/A l 8. Credit for Rad ==c'ive Decay in No No Transit to Dose Point

9. Ims of Offsite Power Yes Yes 4

4

                                                                 =  r-l C.           AmW Dispersion Data                     <

Table 2.3-1 1 Tabic 2.3-1 f D. Dose Data k lb ! 1. Method of Dose raiminnaa A,Wir 15A Appendu 15A ' ]

2. Dose Conversion Assumpoons AWir 15A x Appendu 15A i

C V

6 9,V
& <yP- l wA .

es en - y\M g ar ) q ~'v 4 1 4 h Amoreweef Doesyn Afessyw. Acaslept A@ Pope f5.f JJ

3 J i l System 80+ Deslan Conem! Document i i Table 15.1.5-13 Radiological Consequences of Steam Line Breaks Outside - Containment Upstream of MSIV i i 3 ! Offske Doses (rem) I i Location SLBFPLOPD (Case 5) SLBZPLOPD (Case 6) i

1. Exclusion Area Boundary (0-2 hours)

a. nyroid 70 28' (PIS) g 2

b. Whole-Body 0.3 0.06 (PIS) 0.06 (GIS)

. 2. Im Populanon Zone (0-8 hours) l l

i a. n yroid 34 9.5 (PIS) j 20.6 (GIS) l i b. Whole-Body 0.08 0.014 (PIS) 1 i 0.016 (GIS) 1 1 1 1 i I I i J 1 1 Values are provided for a Pre-existing lodine Spike (PIS) and for an event Generated lodine Spike (GIS). w oemp noeauw Aeemane Annpeer noe 1s. t.u 4

a a. System 80+ Deelen Control Docsunent Table 15.2.3-2 Assumed Initial Conditions for LOCV Pamneter Vahm initial Core Power level, M/ 3992 X Core Inlet Coolant Temperature. *F 550 6 Core Mass Flow,10 lbm/hr 154.3 Pressurizer Pressure, psia 2250 Initial Pressunser Water Level, percent of wide range 26 Initial Core Mimamm DNBR 1.63 Integrased Radaal Pealung Factor 1.53 Steam Generasor Water Level, percent of wide range 76.4 CEA Worth for Trip,10 2 Aa -8.86 d 0.0 Moderssor Temperanne Coefficient,10 Aa/'F Doppler Reactivity Table 15.0-5 Table 15.2.8-1 Assumptions for the Lhniting Case Feedwater Line Break Event Parameter Assumed Yahme Initial Core Power, MWt 3992 Initial Core inlet Temperamre, 'F 561 Initial Core Mass Flow Rate,10' lbm/hr 152.0 Initial Pressunser Pressure, psia 2275 Fuel Gas Gap Heat Transfer Coefficient, 450 Btu /Hr-ft 2 .p Doppler Coefficient Multiplier 0.85 Pressunzer Safety Valves Rated Flow Rate per 525000 Valve, Ibm /hr laitial Pressunser Liquad Volume, ft 3 1400 Initial Steam Genensor invensory, Ibn 170000 Initial Feedwater Enthalpy. Btu /lba 376 Steam Bypass Control System Manual Normal Onsite or Offsite Electncal Power After Unavailable Turbine Trip Feedwater Pipe Break Area, ft2 0.6 CEA Worth at Trip,10 2 as -8.86 d Moderator Temperature Coefficient.10 Ao/*F 0.0 Doppler Reactivity Table 15.0-5 Aweved Deepr asenennt Acedent Ane&een Page r6.2.r5

i j aL i ) System 80+ oenlan ControlDocument i i Table 15,3,3-1 Sequence of Events for the Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting from Turbine Trip l l j Setpoint ) 4 i Time (sec) Event Seapoint or Value j 1

 !                 0.0      Seizure of a Single Reactor Coolant Pump                                        -

0.22 Low Reactor Coolant Flow Trip Condition Reached, Fraction of 0.80 ) Hot Leg Flow j 1.27 14w Reactor Coolant Flow Reactor Trip Signal Generated - 1.42 Reactor Trip Breakers Open - l 1.42 Turbine Trip / Generator Trip 1.42 Loss of Offsite Power Occurs f j 1.92 CEAs Begin to Drop into the Core - l 3.40 Mini =nn Transient DNBR 1.09 { ! 4.88 P essuruer Safety Valve Open, psia 2540 Mani==n RCS Pressure, psia 2615 ) 5.01 11.20 Pressunser Safety Valve Close, psia 2070 j 2 11.80 Main Steam Safety Valves Open Unaffected leop, psia 1212 12.70 Main Steam Safety Valves Open. Affected Imop, psia 1212 19.2 Maninnun Steam Generator Pressure, Affemed Loop, psia 1248 20.1 Maninnun Steam Generator Pressure, Unaffected Leop, psia 1247 545.7 Steam Generator Water Level Reaches Emergency Feedw 19.9 , Acomian Signal Analysts Seepoint in the Unaffected loop, Wide Range l ] 605.7 E y Feedwater Begas Entenns Steam Generator, 68.7 i Unalfacted Loop, Ibm /sec ! 1224.6 Steam Generator Waser Level Reaches Emergency F 19.9 l #ainaan Signal Analysts Seepost in the Affected Leop, 5 Wide Range j 1284.6 Ei-. i Feedwater Begas Ensering the Steam Genermor, 68.7 j Affected Imop, Ibm /sec I l 1784.6 Main Steam Safety Valves Close, Affected and Unaffected I. cop, 1151 l Psia l 1800 Atmospheric Dump Valves Opened to tamase Plant Cooldown, -100.0

                             *F/bour One Annampheric Dump Valve Sdeks Open l

l 3600 Block Val've Upstream of Stuck Open ADV Closed - i j 28800 Shutdown Cooling initiated, RCS Pressure /Temperamre, psigF 330/350 I \ l [%s.(*F 1 i ) W Deep AfeferiW. AcaWom Anahses pops yg.17 7 4.____.

... - g - - . - . _ - -_ . _ _ . - .--- - - - - .- --- _ i i i l System 80+ Denkn ControlDocurnent ; ! Table 15.3.3-2 Assumptions and Initial Conditions for the Analysis of Single Reactor l Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting i from Turbine Trip

1 f

Paranneter Value l l i Core Power 1.svel, MWt 3992 l , j Core Inlet Coolant Temperature. 'F $50 ; k Reactor Coolant System Pressure, psia 2337 l Pressuriser Pressure, psia 2325 l

Steam Generator Pressure, psia 963 Core Mass Flow.10' lbm/hr 154.5 1 Mamnum Radial Power Peaking Factor 1.66

Core Minunen DN3R 1.62 Doppler Coefficaent Multipher 0.85 CEA Worth for Trip,104 4 (most reactive CEA Stuck out) -8.86 Moderssor Temperamre Coefficient,4x10d/ *F -0.1

 !              Doppler Reactivity                                                                                 Table 15.0-5 i            f'  A.        Data and Assumpoons Used to Evaluate the Event's Radianerive Source Term l                           1.        Core Power level, Mwt                                                               3992

{ 2. Burnup, MWD /T 28,000 l 3. Percent of Fuel r nientsand to Exponence DNB, % 1.2* i 4. Reactor Coolant Actmey Befose Event Appendix 15A ! 5. heandary System Actmty Before Event, pCi/gm Appendix 15A l 6. Prunary Symem Lkpud. " . y,1bsn 605,000 { 7. Steam Generator 1. y j - IJquad,Ibn per steam generssor 197,000

 ,                                   - Steam, iba per steam generator                                                   15,160 I          I     B.        Data and A===paa== Used to Esnesse Acamty Released from the knandary                                          b I                 System i                           1.        Primary-to-Secondary Laak Rase, gym                                             1.0 (total)

. 2. Total Mass Release Through the main Steam Safety Valves, Ibn Table 15.3.3-4 l 3. Total Mass Release Through the ADVs from 30 to 120 Minutes,Ibn Table 15.3.3-4 j g 4. Percent of Core Fieman Products Asemned Released to Reactor Coolant See App.15A i j 5. ladine n=ca====ia= dan Factor for the Unaffected Steam Generator 100* l W 6. loshes n.nnne===== dan Factor for the Affected Steam Generasor 100* j 7. Craht for Radaoactive Decay in Transit to Dose Point No i 8. Less of Offene Power Yes i C. Ati- - p: Dispersion Factors j 1. at EAB,0-2 hr, sec/m3 Table 2.3-1 1 2. at 1.PZ, 0-8 hr, sec/m3 Table 2.3-1

  !             D.        Dose Data i                          1.       Method of Dose Calculation                                                   Appendix 15A

) ( 2. Dose Conversion Assumptions Apindix 15A

                                                                                                                                     )l i

l l l l Anwevoet Dee@e Aseannd Aeomant Annween Page r5.2.r2 l

c _ 4 System 80+ oesten control oocument Table 15.3.3-3 Parameters used in Evaluating the Radiological Consequences of a Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting from Turbine Trip Parameters Valm A. Data and Assumptions Used to Evaluate the Event's Radioactive Source Term

1. Core Power level, MWt 3992

2. Burnup, MWDfr 28,000

3. Percent of Fuel Calculated to Experience DNB, % 1.2 *

4. Reactor Coolant Activity Before Event Appendix 15A

5. W='y System Activity Before Event, pC1/gm Appendtx 15A

6. Prunary System Liqmd Inventory, Ibm 605,000 7.

ig Generator inventory: g 7QMoT .Liqmd, Ibn per steam generator 197,000 h Ibm per steam genermor 15,160 N B. Data and Assumpuons Used to Farimme Activity Released from the Wa y System

1. Pnmary-to hwvimy Laak Rate, gym 1.0 (total)

2. Total Mass Release Through the Main Steam Safety Valves, Ibn Table 15.3.3 4

3. Total Mass Release Through the from 30 to 120 Minutes, Ibm Table 15.3.3 4'

4. Percent of Core Fission FE t= n.s., =n=M Released to Reactor See App.15A K, Coolant

5. Iodine naron. ..v= Factor for the naffected Steam Generator 100"

6. Iodine naeaar===h Factor Steam Generator 100 "

7. Credit for PMeve Decay ransit to Dose Point No

8. Ims of Offsite Power Yes C. A@ Dispersion Factors

1. at EAB,0-2 hr, sec/m3 Table 2.3-1

2. at LPZ,0-8 hr, sec/m 3 Table 2.31 D. Dose Data

1. Method of Doesc'etent==a Appendix 15A

2. Dose Conversion A**'=Taaae Appendix 15A a

1

MLCT1

             " Tun ceLuas LLok No failed fuel assumed for GIS doses.

Also, applicable to Cesium and Rubidium for the failed fuel case (see Appendix 15A). Anreved w nennenw Acewent Annopean rene rs.2.r2 j

e System 80+ Desinn Conmd Document Table 15.3.3-4 Secondary System Mass Release to the Atmosphere for the Single Reactor Coolant Pump Rotor Seizure with Loss of Offsite Power Resulting from Turbine Trip Evpdf/ /,e h/j r - __ / ~:- , -- e,w .: - '

                                                          -1 km           ,w        :    so u idw \                         K Time (sec)         Unaffected        )   A fecteT      )    Lineff                                     laci.

0.000 197021. 11.200 205082. 74.100 180933, 263.60 157739. 407.40 146831. 553.00 138161. 682.80 145915. 848.60 140345. 1038.2 140420. 1229.8 140556. 1503.2 139849. 1784.6* 140593. 1800.0** 141528. 3600.0 141528. 7200.0 141528. 28800.0 141528. Table 15.3.3-5 (0-2 hours) Thyroid l@* 4TM had Whole-body J W'M Thyroid hl Qutt,g.M ' 1.33 197021. O. O. O. 219448. O. O. O. 212331. 25977. 17918. 43895. 202082. 57455. 17918, 75373. 192151. 67292. 27689. 94981. 181875. 76411. 36938. 113349. 172793. 76411. 47032. 123443. 161723. 85395. 56016. 141412. 149485. 94118. 64804. 158922. 138010. 102579. 73265. 175844. 139765. 111039. 81725. 192764. 139864. I19367. 90053. 209420. 140676. 119367. 90053. 209420. 140676. 119367. 90053. 495800. 140676. 119367. 90053. 980900. 140676. 119367. 90053. 1961500. Radiological Consequences of a Pa=*ntatd Single Beartar Coolant Pump Rotor Seizure with Imss of Offsite Power Resulting from Turbine Trip 14cesion Deses inna seemedary symem steen Reissess, rem Exclusion Ana Boundary $9 c.a.uT6t, GIS Failed Fuel I K g 0.6 3.18 4 0.02 0.13 low Populanon Zees 5 (0 8 hours) 2.18 X Whole-body I 0.01 0.04

                                                                                                       /
                                                            ~

QQO % Wen t" -. l NsSVI WokcO Main steam safety valves close. O_naEds45Gi/keelG D^d 4 - - l

                                                            \                         \

Operator begins cooldown utilizing the atmospheric dump valves. One atmospheric dump dalve is assumed to stick open for the next 30 minutes. Anwreved Deep Meeuw Acoment Aneween rose rs.314

, r_ _ ] A 2, System 80 + Design ControlDocument ! whem: M = RCS mass C = time dependent RCS boron concentration W = Charging mass flow rate of unborated water dC/dt is marimi'M by marimi'ing W and minimi'ing M. Assummg: W = Constant, equal to the maximum possible value, and choosing: M = Constant, equal to the mmimum value occumng durmg the boron dilution incident, the solution of Equation (1) can be written 4---

                                                       ~%N (C, - C;'*                 C = C, e.,                                                          (2) wiw:e:

r = M/W = Boron dilution time constant Co

                        =       Initial boron concentration The time T required to dilute to critically is given by T        =         rin (C,/C.)                                                                (3) where:                                                                                            i O

C. = Critical boron concentration l 0 e Input Parameters and Initial Conditions it is assumed that the inadvertent deboration proceeds at the maruman possible rate. For this to occur, the charging pump must be on, the reactor makeup water tank must be aligned with the charging pump suction, a reactor makeup water pump must be on, letdown flow must be diverted n from the volume control tank, and a failure in the boric acid makeup water flow path (e.g., flow control valve, CH-210Y failing in the closed position) must tenninate borated water flow to the

charging pump suction. j w Deeen neeeuw. Aeament w pope yg.s.1o I

;_     = 5 : ; .::_            z ..: -       = - -- - - - -

s - a .. o i i l Svartem 80+ Danton ConeelDocunnent i 1he tr-ient behavior of the NSSS followmg a p~ad"M CEA Ejecuon is as follows. Tlie steam generator pnssure meresses rapidly due to the closme of the turbine control valve j followmg reactor and turbine trip. The steam bypass control system is inoperable on loss of offsite power and therefore is unavailable. The steam generator pressure reac1== a vaarinnan of , l I i 1279 psia at 8.05 se.xmds. The pressurizer pressure increases to a snarinnan of 2399 psia at 5.4 seconds due to the decreased heat removal of the steam generators. i Subsequently, the reduced reactor power following the reactor trip causes the RCS pressure and l i tesapersmre to decrease. i ! 'Ihe steam generator pressure fhusnates as the MSSVs open and close throughout the event The i total steam release through the safety valves is apprari-+-ly 317,100 Ibm up to'1'800 seconds. t Following a panenlarad CEA ejection event, 6.8% of tid fuel is calm 1=*M to experience DNB. l Regulatory Guide 1.77 rarananande that the onset of DNB be used as the basis for predicting clad faihne. It is conservanvely ====nad that all fuel rods experiencmg DNB suflier clad faGue. A separate analysis was N----i to enarimiw the RCS pressee. The initial pnoseriner f Pressure for this case was 2400 psia. 1he peak RCS pressure reaches 2742 psia (including the j 4 pressure difference between cold leg at the RCP hs. and the surge linch This value is less than 120% of design pressure. Figure 15.4.8-15 shows the peak pressure transient Radiological consequences for this evem are cal =Israd for two cases: release thmugh

                   """"=nr and release through prunaryWry leakage.                                                     _ , , ,

3 l The ammmnprians, parameters and e=lenimianal ==hade used to evaluate the radiological i consequences of Chapter 15 events are diaenenad in Anp==dir 15A. Assumptions and parameters l , that were unique to the evaluation of a CEA Ejection Event are itemized in Table 15.4.8-3; The j following paragraphs provide additional clarification to some of the items containad in the table.

1. Actmty avadable for release from mnramment at time zero

! 1he activity available for leakage frorn marminmane is based on the followmg Regulatory ! Ouide 1.77, App--dir B assumption as modified by Draft NUREG-1465 for the gap ! acamty: ! 1he acavity in the fuel clad gap is assumed to be 5% of the indinaa,5%'of the noble gases and 5 % of the canimn and rubidium amannlarad in the fuel at the erzi of core life, l assummg continuous .. .... ... full power operation. AH of the activity 1.i tinfuel gap for fuel rods that are calmia'M to experience DNB is assumed to be '- --- - --- +f mixed C.@ ennrainmant and avadable for leakage to the 4s. . l.

2. Activity Release from the W=y System g . )%" ]

] Actmty released from the *M 2- by system based upon the =+- 'E f activity initially f resulting from technical specification steam X M1 in the steam generators plus prunary generator tube leakage. Table 15.4.8-4 con activitzhins the total steam safety valves and the total prunary to secondary leakage. The mass of steam mass releases frorn i released through the ADVs is given in Table 15.4.8-3. i i 5 Aaproved Deser asseenw - As= Mast An=Wes her ff.4 78 u

                                                                 ;_       _ _ z                 -

< ;p- -., .s g. . r . .., . . a Si -1 ;; 80 +- Damian consruf Doessment f, t , t

                                                                                                                           "* d

! 15.4.8.4 t'ancammians ' i 11ie ejecnon of a CEA wi'dnot result in a rachal average fuel enthalpy greater 280 cal / gram at any / sysaun steem j, axial location in any fbal rod. The rachological consequences associated l releases have been m-,ay analyzed usag assumptions and models desenbed in the preceding 3 e j % secnons.11ie#4-body dose due to immersion and the thyroid does due to inhalatian have been j analyzed for the two-hour dose at the ereinaian area boundary and are presented in Table 15.4.8-5. The j 30 day doses at the low population zone have been found to be less limiting than the EAB doses. 11 ire j resultam doses are less than the allowable site be==wlary dose set forth in 10 CFR 100. Control room { doses are provided in Section 6.4.3.- , d j 1 Tire peak RCS pressure for the CEA Ejecnon event is 2742 psia. 11ds is less than Service Limit C vaine as W in the,AShtE Code.

I g e

ll i 4 0 j 1 i. j l v - : . - rg. ) . _ . . . . . . _ . . I I B i i , i i e l 4 ( i 1 1 4 i . . i l l ) i Asymes pos(pe messment. AseWmW Asenfpose AUS fe 4*II i . . ~-

 .7.--

f l 4 2. I l System 80+ Design ControlDeewnent ! i . l I Table 15.4.8-3 Paratneters Used in Evaluating the Radiological Consequences of a l CEA Ejection Event Parameter Value A. Data and Assumptions Used to Evaluate the event's Radioactive SourgTerm %

1. General

a. Core Power Level, MWt 3992

b. Burnup, MWD /MT 28,000

c. Percent of Fuel calailead to Exponence DNB, % 6.8

d. Percent of Fuel calailmad to Expenenz 'T 0.0 Centerijne Melt, %

e. Reactor Coolant Activity Before Event Tech

f. Secondary Syssen Activity Before Event Tech B

g. Primary System L.iged Invensory, Ibni 605,000

h. Steam Generasor lavsesory

                                    -        1.iged, Ibna per steam generaser                      117,000 Steam, Ibni per steam generasor                       23,700 l

i. Average peaking facsor 1.3 I

B. Dasa and A====raaan Used to Estunsee Acuvity a.i d l

1. canta= =am tmakage

a. raara==== Vohans, ft3 3.34 E06

b. Fantain= mar IAak Rase, vol. 5/dey Section 15.6.5

c. Percent of Core Fission Products Anemned Released to Refer to Secuan Paar====ar 15.4.8.3

d. Namral Deposition inraarnia===r Yes i A = 0.15 hr 1 l for paruculate A = 2.89 hr 1 for elemental iodine

e. Credit for Radioactive Decay Hold up in Contam=ent Yes In Transit to Dose Point No Anwend DeeQn aseaud = AseMant Ane%ee Page r$.4-M

t^ U i System 80+ oesian comrolDocument

;           Table 15.4.8-5            Radiological Consequences of a Postulated CEA Ejection Event I

Thyroid Dose Whole-Body Dose ) Release Pathway (reen) (nan) l i Exclusion Area Boundah-hour consequences r Via Containrnent 69.8 0.7

   #    L            Via w ry                                 17                         0.6
;     )

j Iow Popn'=iaa Zong 30-day consequences

!  #Mf-              Via Coneninenent                        14.5                        0.2 I

cell - Via Secondary 11.3 0.2 3 4 4 i 1 J -f i i i l l 1 + 4 9 AnwwM wap Anew Acainant Anarpeen pop, ys,s.27

i ,A L i System 80+ Deelen ControlDocument l _ l generator tube rupture (SGTR) event, and (2) as shown in Section 15.6.3.2 for an SGTR event j with a loss of offsite power, the muumum DNBR remains above the SAFDL of 1.24. l l A Letdown Line Break event in combination with a loss of offsite power is less limiting.

Therefore, the final analysis did not assume a loss of offsite power.

15.6.2.3.2 Radiological Conseqi=ne== } e Mathematical Model j De DBLLOCUS event is indicated by several alarms listed in Table 15.6.2-1. Thirty mimitaa j after the first alarm, which takes place immediately following the initianon of the event, the ! leadown line is isolated by the reactor operator. Durmg this time 48,617 poumis of primary coolant is released to the nuclear annex. } i l The methodology used to calculate the inhaintion doses at the Exclusion Area Boundary (EAB) j and the law Population Zone (LPZ) is discussed in Appmiir 15A. e Assumptions and Parameters l The letdown line break outside cantamment results in the h of radioactivity to the j environment. Dere are some uncertamrias in the calculation of resultant radiation doses. Thess

principally arise from uncertaindes in the reactor coolant activity levels, the quantity of coolant i released, the fraction of radionuclides that become airborne, the fraction of airborne activity that escapes the nuclear annex, and meteorological conditions that exists at the time of the accident.

4 These uncertamties are treated by taking worst case or conservative assumptions. These are: l . An iodme ac,:ivity spike with a spiking factor of 500 is assumed to occur coincident with j the initiation of the transient. I

2. The quantity of coolant released outside containmant is maximimi by assumag most

! adverse initial conditions and by assumng crincal flow through the break. i

3. A blowdown Darantammarian Factor (DF) of 5.05 is assumed in the calculation. That is,19.8% of all the iodine contained in the released prunary mass is maanmad to be sirborne. This is based on the fraction of prunary fluid that flashes to steam in the nuclear annex based on the enthaply of the escaping fluid.

4. The nuclear annex DF is assumed to be 1. That is, no credit is taken for the retention or filtration of radioactivity.

5. No credit is taken for ground deposition of the activity that escapes the nuclear annex or of decay in transit to the exclusion area boundary.

6. Other assumptions are contamed in Appendix 15A.

e Results During the 1800-second duration of the transient no more than 48,617 pounds of prunary system coolant is released outside the containment. The total secondary steam mass release via the Anumn oweeeenw Aumn Anawee rene rs.s.4

System 80 + Deslan ControlDocument i ! system pressure until the pressurizer begins to refill at approximately 5583 seconds. Due to ! flashing caused by the depressurization, and the boil off due to the metal structure to coolant heat transfer, the reactor vessel upper head begins to void at about 1794 seconds (Figure 15.6.3.38). { l Consequently, the RCS pressure (Figure 15.6.3 34) begins to decrease at a lower rate at this time. ! Following reactor trip and with turbine bypass unavailable, the main steam system pressure i increases until the MSSVs open at 1762 seconds to control the main steam system pressure. A j maximum main steam system pressure of 1272 psia occurs at 1765 seconds. Subsequent to this i peak in the pressure, the main steam system pressure decreases, resulting in the closure of the ] Main Steam Safety Valves (MSSVs) at 1849 seconds. De MSSVs cycle two additional times i in this manner prior to the operator takmg control of the plant. ) Prior to reactor trip, the main feedwater control system is assumed to supply foodwater to match flow through the turbine. Following reactor trip, the main feedwater flow is terminated due to the loss of offsite power. As the level in the steam generators decrease an Emergency Feedwater l Actuation Signal (EFAS) is generated resulting in, emergency feedwater flow which acts to j restore the SG level. . g J J j At 2178 seconds the operator takes control of plant and opens one ADV on each SG to cool i down the plant. This is comistent with the . At 3663 seconds the RCS has been cooled { to 550*F. The operator then isolates the emergency feedwater to the affected generator, closes j

j. the main steam isolation valves of both steam generators, and attempts to close the ADV of the 3 affected generator. The operator recognizes that the ADV did not close and has the appropriate block valve closed within 30 mimum. The operator then initiates an orderly cooldown by means

{

of the .i-s.$wic dump valves and the emergency feedwater flow to the unaffected steam

! generator. Thereafter, the operator will steam the affec+ed steam generator only for preventing

overfilling due to the leak flow. AAer the pressure and temperature are reduced to 330 psia and

( 350*F, respectively, the operator activates the shutdown cooling system and isolates the 2 unaffected steam generator. , ne potential for SG overfalling exists due to the contmswi break flow from the prunary side to l the ="+ 4= y side of the affected SG even aAer isolation of this SG. However, the System 80+ ! design can accommadare the SG level bmidup more easily due to the larger SG volume in l

comparison to that for the System 80 design (about 25 % more volume).

1 ! Durmg the SGTR event, the operator momtors the SG level to ensure that the level does not go i much above the steam separators (around the upper level indication nozzle). This is j accomplished by (1) mmimmng the differennal pressure between the prunary and secondary sides 4 of the steam generator (to control the break flow) and (2) by selectively steaming the affected SG via one of its two ADVs. ] Figure 15.6.3-42B shows that, after about 15000 seconds, the break flow rate drops from

approximately 35 to 12 lbm/sec. At this time, the operator is expected to use SG level control j to prevent potential SG overfilling. Based on critical flow of steam through one ADV at the i steam pressure, a flow rate of about 70 lbm/sec can be achieved. His suggests that the break l flow rate of 12 to 35 lbm/sec can be accommodated by partial opemng of one ADV (17 to 50%)

{ to maintain an essentially stable steam generator level. l i Anwevent Deeen acnonlah AceMont Anerpoen Pope 1e.e.19 }

ar N

System 80+ Deslan ControlDoctamtt the greater lambdas used in the PAG comparison analysis is that these lambdas include the effects of hygroscopicity. The hygroscopicity of the CsOH released from the prunary system into the Mm saturated contamment atmosphere increases the airborne particle sizp. distribution and maken the Q-7 sprays more effective.

     }e                                                                           9- @ a"- 4oWh(.1M(a3f
 /          PAG offsite dose calculation is done with the MACCS computer c+f ';; ^, r C fv.

14ACCSC J.-uyuvu)Tcording to the following set of assumptions:

The reported dose is the Commined Effective Dose Equivalent (CEDE) since this is the way in which the revised (October 1991) EPA PAGs are specified.
A median dose is reported indWant of direction. Since the event itself represents a worst-case ,

severe ==ient (given an intact conenmment and sprays +=ing), the reporting of the madian l i dose, i% of direction, establishes that most severe accidents with an intact containment and sprays operstmg will result in lower offsite doses.

The meteorological data base is the same as that used to generate the X/Q values used for the DBA LOCA analysis.
50 year dose conumtment is used.
The following values of "a" and "b" are used for the expression sigma, = axb; i

Stabihty a h A 2.47E 4 2.118 B 0.078 1.085 l C 0.144 0.911 l D 0.368 0.6764 l E 0.2517 0.6720 F 0.184 0.6546 i

The time base for plume maander is three mimmen with a 24-hour release duration.

i e The release is assumed to be a cold, ground level release. i

Doses include a 24-hour exposure to ground contammation.

1 I l

Doses include shielding factors of 0.75 for plume exposure and 0.33 for ground exposure. 1
The breathing rate is 3.3 x 10d m3 /second with an inhalation protection factor of 0.4.
The dry deposition velocity is 1.0 cm/sec for iodine and 0.1 cm/sec for other particulates.

l 1 The results of this analysis are as follows: 1 ! Lumi Dose Result ! CEDE PAG 1 rem 0.33 rem l Thyroid PAG 5 rem 2.7 rem Anmed coayr anesener. Acouerre Anemm n oe rs.s. n

   . . . - _ - . . . . . .                   - . -               ..-_- ~ . .          .- . . . .     -          . --         -. .-.

i l System 80+ Declan ControlDocunwrt Table 15,6,3-1 Sequence of Events for the Steam Generator Tube Rupture i Time (Sec) Event Setpoint or Value

0.0 Tube Rupture Occurs -

0.4 High Steam Generator Level Trip Signal Generated - 0.55 Trip Breakers Open 0.55 Turbine Trip: Stor Valves Start to Close -- 5.4 Main Steam Safety Valves Open, psia 1212 5.75 Main Steam and Feedwater isolation Valves Closed 8.77 Marinum Steam Generssor Pressure, psia 1273 16.5 2325 Backup Hessers _T.nergjis , psia 623 Pressunser Heats Deerargize due to Law Pressunser Liquid 297 Volume, ft3 1800 Operssor Isolases the Damaged Steam Genermor and initiates Plant - Cooldown at 100*F/hr for the 1.5 hour thne pened 28,800 Simadown Cooling Entry c'aadman are Assened to be , 330/350 RCS Pressure, peia / RCS Temperamre, 'F p Table 15.6.3-2 Assumptions and Initial Conditions for the Generator Tube Rapture Parameters Assumed Yahse Core Power Level, MWt 3876 Core laict Coolant Temperamre, 'F 563 Pressunasr Pressee, psia 2375 Core Mass Flow Rase,108 lbadhr 151.9 One Pin lategnsed Radial Peaksag Factor, with Uncenaary 1.46 Steam Generasor Pressure, psia 1057 Moderator Temperature Coefficient,104 bl*F 0.0 Doppler Coefficient Multiplier 1.0 CEA Worth at Trip, % 4 (most reactive CEA fully withdrawn) -8.86 Doppler Reactivity Function See Table 15.0-6 Approved W Asesanimf Asedener Anafpees pope rf.5-30

i N1 System 80+ Dealers coneet Document Table 15.6.3 3 Radiological Consequences

of the Steam Generater Tube Rupture Thyroid Inhaintion Doses Thyroid fahalmeta= Doses y Offsite Doses" (ress)

I4 cation GIS PIS h o Exclusion Area Boundary 0-2 hr, Thyroid 7.05 h 27 .6

                      ,,.sL_                     Iow Populanon Zoos                                    3.25                4 Outer Boundary                              {9                  h .48 0-8 hr, Thyroid Whole Body Dooms y                        Ofhite Doses ** (rum)

IAcaties GIS FIS QAj Exclusion Area Boundary (0.07 /0 0-2 hr. Whole-Body \J. h .09

                   . ,y                         Low Perth Zone                                       0.015
                        -                               Outer Boundary 04 hr. Whole-Body (4 0.016 Radiological consequences were deternuned for a core power of 3992 MWt by increasing the steam releases for the 3876 MWt core power case by 3%

GIS - Generated Iodine Spike PIS - Pre-accident fodine Spike

               & De*** M A m W                                                                                            rene ss.s.21

gs System 80+ oesion control Docenent Table 15.6.3-5 Assuinptions and Initial Conditions for the Steam Generator Tube i Rupture with a IAss of Offsite Power Parameter Assumed Value Core Power 1.evel. MWt 3876 ' Core inlet Coolant Temperature, 'F 563 Pressurizer Pressure, psia 2375 l Core Mass Flow Rate,106 lbm/hr 151.9 l One Pin Integrated Radsal Peakie.g Factor, with Uncertamty 1.46 Steam Generator Pressure, psia 1057 l Moderator Temperature Coefficient,10 Aa/*F d 0.0 ) j , l Doppler Coefficient Multiplier 1.0 CEA Worth at Trip. % Aa (most reactive CEA fully withdrawn) -8.86 l . Doppict Reactivity Faad*lr Funcuan See Table 15.04 l 1 I j Table 15.6.3-6 Radiological Consequenra.* of the Steam Generator Tube Ruptum j with a IAes of Offsite Power i Hyroid Imbdanna Domes l ] { Imentical y Ofhite Deses" (rem) GIS FIS i I 9 38 .6 fl." Elxc uhm" Area Boun:EyD l >

0-2 hr, nyroid ( --e. 6A b .53 b

2. Iow Populauca Zoos 4.23 5

/'

A Outer Ranadary q.8 hr m n yroi o ( h .98 i Whole Bedy Doons IAeStioE y Olhite Doess" (rous) l GIS FIS

l [l'.' *'Excluson Aree Boundary 0.112 0.143 j i 0-2 hr. Whole-Body d I l L2. Iow Para h Zone 0.019 0 1 L u Omer Boundary D 8 hr, Whole __ h .02 t ] i Radiological releases were detenntned for a core power of 3992 MWt by increasing the steam releases for ~ the 3876 MWt core power case by 3% I GIS . Generated Iodine Spike P!S - Pre-accident lodine Spike 1 komed Deepr Asenennt Aconnent Anarpoon Page 15.6 33 i

. a. L. 1 I i System 90+ Design Conmd Document

Tabk 15.6.3-8 A- p9ns with Initial Conditions for the Stearn Generator Tube Rupture with a Loss of Offsite Power and Stuck Open ADV Parameter Aasaned Value j Core Power Level. MWt 3876 Core Inlet Coolant Tswi.iws. 'F 563 j Pressurizer Pressure, psia 2375 Core Mass Flow Rate 10'lbm/br 151.9 One Pin Integrated Radial Peaking Factor, with Uncenannty 1.46 j Steam Generator Pressure, psia 1057 d

3 Moderator Temperature Coemeient.10 b/*F 0.0

;                Doppler Coemeient Multiplier                                                          1.0 j                 CEA Worth at Trip % b (most reactive CEA fully withdrawn)                             3.86 l                Doppler Reactivity FWhacir Funcuon                                              See Table 15.0-6 a

Table 15.6.3-9 Radiological Consequences

of the Steam Generator Tube Rupture with a I4es of Offsite Power and Stuck Open ADV Dyroid inhanneiam Doses j y OtMee Doses" (rums) j i ar=ela= GIS FIS J

1. " Exclusion Area -

                     "^-                                         C4uML                                       93.1) 0-2hr,~1hyroid                                          62.4 )

{ . Iow Pcpulanon Zone 47.9 30.9

                     .   ,         Outer Boundary j                       % J hr. Ihyroid Wr:        fD

) y OtMee Deses" (rous) ] I me=ela= GIS PIS l 6. - Exclusion Area Boundary 7.42 7.5

0.2 hr. Whole-Body -- 4

. Qoff&

2. Low Parnimaa= Zone 1.082 1.061 A Outerar==tary L __ 0 8 hr. Whole-Bod _v l

Radiological consequenm were determmed for a core power of 3992 MWt by increasing t!r: steam releases for the 3876 MWt core power case by 3% i ** , GIS - Event Generated lodine Spike l PIS - Pre-Accident lodine Spike Approwd Deep Afeennel- AccWorn Ansepoos Pope r5.5.75

i' C- - wL System 80+ oestan comrol Document Table 15.6.5-2 Parameters used in Evaluating the Radiological Consequences of a l I4ss of Coolant Accident (LOCA) Parameter Value

  .Jd a. 1.  , Core Data
                 ' Power Level MW:                                                                         3992 Ate la.                                                                                                 28,000 M              Core Average Burnup MWD /MTU

2. Accident Releases '

The radioactivity released followmg a LOCA are from the following sources: j - Blowdown of pnmary coolant

                  -        100% of the gap activity l
                   -       Fracuan of the Core Activity due to melted fuel l

Blowdown of Pnmary Coolant Primary Coolant Techmcal S, 1"-~='= enneenr r cas

                                                               -                                        Table 15A-4 Pnmary coolant activity with a Pre Accident lodiac spib                      Table 15A 6 Duranon of RCS blowdown via break, secs                                           20 Normal Pnmary coolant mass, iba                                                630,198 RCS Flash Fraction, 5                                                              54 Power Purge Rate, cfm                                                           16000 Purge Duranon, secs                                                               30              <

Credit taken for raarmammat low purge filter No Chenncal species of the iodines in RCS (i.e. a I-spike release) > Gap Release Gap activity released (based on 5% of the core Noble Gases, iodmes, Cs & Rb) Table 15A-3 Cbeancel species of iodines in the gap, 5

                           -        Inorgame i ===ei                                                              4.75 U ler .                 Peruculme                                                               95 2tws*      (                                                                                       )

SNat . Organic 0. Duration of gap reless, secs 1 5 44, reisese nans, secs *

                     ** W             release ends, secs                                                    1830       *F An"*M Deep aseaud Aeenant Ane%een                                                               pop gg,g.27

System 80+ Deslan ControlDocument C,(i) = C,(i)/D, t EC i 3 = MPE(i)/D,

     ~                                      A re:

C,(i) = Concentration of the i-th isotope in the tank (pCi/ml) C,(i) = Concentration of the i-th isotope at the nearest potable water supply i/ml) , , 3 ~ 60 \ E f kk 3 @1 -(i) = k'MRuant' M

                                                       "-     "-LConcentration of the i-th isotope (pCi/ml)
                                                                     ^
                                                                           ^

n so ., ~ , . - t Dr = Dilution factor b e , e N g?, ' . y ,~ ,

(1/ Dilution Volume)(1/ft')h0. *iV, (gal') Y '. / enConversion Factor (ft'/ gal) ,. / . j

s. . ,

V, = Volume of liquid ' initially in the tank (gal) d,. , i The concentration in the liquid flowing into the Boric Acid Concentrator from the Holdup Tank l is calculated by utilizing the following equations: C,,fi) = C,(i)/[DF(i)@DFfi)@... i DFfi)) y i e: 9 g] C n(i) i

                                =         Concentration of the i-th isotope in the holdup tank ( Ci/ml)

C,(i) = Cn--arration of the i-th isotope in the composite flow stream (pCi/ml) as a function of PCC DFfi) = DF of the j-th component in the purification process for the i-th isotope C, is calculated as follows: 3 3 X C, = E(F3x C)/E(F) 3

i X re: ,

F3 = Flow Rate of the J-th flow stream (gpm) Cj = Concentration of the i-th isotope in the j-th flow stream (pCi/ml)

              'Ibe concentration of the fluid in the BAST after the influent fluid from the holdup tank is concentrated is calculated as follows:

C d i) = CF(i) x C,n(i)/DF(i) w Deegpr neesenfel Aeoment Annopees Pnee 15.7 2

              -         - .             . ,    - _             - .                -                          -_.       . _ ~        - _ _ - .

I i i System 80+ Deslan controlDocumart i Where: I j ., CsAsdi) = Concentration of the i-th isotope in the BAST (pCi/ml) , i i j s; CF(i) j = Concentrating Factor between the bottoms and the distillate processed in the j y i Boric Acid Concentrator (BAC) i W = 1.00E+04 (See Section 9.3.4) i i Cudi)l = Concentration of the i h isotope in the Holdup Tank (pCi/ml) 1 l _ l DF(i) ; = Decontammation Factor for i-th isotope. All isotopes are considered to be non-I i volatile; a, nF for the non-volatile isotopes is used. the BAST is shown in Table 15.7.3-1. Then, the rnarmmm l The concentration o allowable dilution a ' o r ( f) is crSufifed. To verify that this is the maximum dilution factor, j the calculated concdn' atri or isotope at the nearest potable water source using the dilution g j em factor is ratioed to

                                                                   .and summed. This process is re dilution factor until the surnmarion of the fractionris i

j 2.g equal to 1.@peated with addition

                                                                                                                                                   't I

Input Parameters and Initial Conditions Ft.C. CQ, v '

                                                                                                           /                      -

l l 1. WO ionm at a nahble water supply is equal to the Manammen K i -$lF

                                         '" T o                                                                                -

24V-+1/ o4(s _, )De.ach isotope. ' j 2. concentration in [eco- ) i is pdated for the batch processing mode of operation j j described in Section 9.3.4 and shown in Figure 9.3.4-1. l ! l I

3. Credit is taken for dilution by only the main flow path from the letdown through the

! purification process in the CVCS to the recycle evaporator. The concentrate is sent to ! the BAST. 1 1 4. 'Ihe concentration of the flow streams are specified as a fraction of the Pnmary Coolant Owatration (PCC). The PCC is obtained from the output of the computer code { DAMSAM for 1% failed fuel fraction (See Section 11.1).

5. No additional credit is taken for radioactive decay during the purification process or 3

dunng the transport of the liquid effluent to the nearest potable water source. j 6. The Decontammation Factors (DF) for the i-th isotope for each component were obtained from NUREG-0017 and Section 9.3.4.

7. System parameters such as flow rates and tank volumes were obtamed from Section Q9 9.3.4.

I i 8. 80% of the volume is assumed to be released per SRP Section 15.7.3. i

9. For conservatism, all radionuclides are assumed to be in the insoluble form.

i Asp =est senes asessaw. AcWdmW AaWpees Aspo 75.7-2 i

l System 80+ Danisn cor;trolDoewnent i

e- Results 1: The results of the iterative process are shown in Table 15.7.3-2. The maximum allowable j dilution factor was determmad to be 2.55 x 104.

15.7.3.4 Conclusions This analysis assumes a emi hvyhic failure of the BAST. Although the BAST is a Seismic Category I i ASME III Class 3 tank equipped with an overflow line, this analysis is considered to envelope any potential failure of a radioactive i ng tank , I for the minimum dilution . i [ (7he resultstooflimitthis flow required the analysis should concentration be used at the nearest towater potable establish e site ec=a= pin t= th=j10_C i j g. Ap-wiir B of Sections 20.1001 - 20.2402, Table 2, Column 2 limits % analysis is based on the once j gh purification of the leQI5Wn~HoliPliiifTquiif from th~e Eq p- = Drain Tank and Reactor Drain i Tank. Also, average values for the concentration of the distillate and concentrate are assumed to ! determme the concentration factor in the Boric Acid Concentrator (BAC). The concentration of the

influent stream to the BAC must be sampled to verify the concentration factor used is valid.

j 15.7.4 Fuel Handling Accident J i 15.7.4.1 Identi5 cation of Event and Cauzs

              'Ile Fuel Handling Accident that is considered resulted from the dropping of a single fuel assembly dnring fuel handling. Interlocks and procedural and administrative controls involved in fuel handling are described in Section 9.1.4.

j 15.7.4.2 Systeens Operation ! The transport of heavy loads in the containment building and the spent fuel building is controlled by j restrictions placed on plant layout and equipment design. Features are selected to ensure that heavy loads

are restricted to preassigned travel zones and that they are not carried over stored fuel assemblies.

Equipment interlocks and prcedes are also used to ensure that load transport is accomplished in a i predictable manner. i l l Containment and fuel building systems required to mitigate the c=~;9m of the Fuel Handling I j Accident are described in Chapters 6 and 9. Restrictions on the cask handling crane are listed in Section i i 15.7.5.2 [ 1

15.7.4.3 Analysis of Effects and Consequences l *- Mathematical Model: If a dropped assembly were damaged to the ext one or more fuel rods were broken, the
accumulated fission gases and iodines in the fuel i gaps would be released to the surrounding i

                        . water. Release of the solid fission products in th: fuel would be negligible because of the low             /

i fuel temperature during refueling. The methods of Regulatory Guide 1.25 were used to quantify 3 the fission product releases to the containment and fuel building. The calculational methods and j .'til g-ih~ activity " -i'h gla-O egulatory4uide 1.25fmodified by Draft NUREG 1465 for the gap ! :f : 'h;-r-?- -t _t since:

1. , col I4mab d44M; .occ TCb JAMurdbt 5+<417 3. L j

u ,- # pops 15.7-4

Svstem 80+ Denian contror Doewnent

             .The chemical specianon and physical form of the iodine released due to the spike are conservanvely assumed to be 100% gaseous; that is, no credit is taken for deposition in particulate fonn once iodine from these sources becomes airborne.

3.0 Dose Analyses 3.1 Assumptions lhe following ==innpriaan are basic to the model for the gamma and beta dose due to immersion in a cloud of radioactivity and the model for the thyroid dose due to inhalation of radianerivity.

                                %                                   r.
               *-       The dose contribunon of direct radiation from sources other than the leakage cloud is negligible compared to the dose due to immersion in the leakage cloud.
               *-       All r-hdvity releases are treated as ground level releaseIregardless of tiie point of ducharge.

e- . The dose receptor is a standard man as defined by the International Camrmazion on Radelogical Protectics (ICRP) (R.,~ w 4). e' Radmacave decay from the point of release to the dose receptor is neglected.

               *-       Isotopic beta and gamma decay energies are taken from Table ofIsotang (Reference 5) and froen References 6 and 7.

e. AP dispersion factors used in the analyses are premmad in Section 2.3.

               *        'Ibe analyses use dose conversion factors based on following guidance daerananen:
                        -        gamma and beta D-deary Guide 1.4 (R            w 8),

thyroid - TID-14844 (Reference 9),

                        -        other organs - EPA 520/1-88 020 (Reference 11).

3.2 Gammaa Dose

               'Ibe gamma dose is obtained by considermg the dose receptor to be immersed in a radioactive cloud which is infinne in all directions above the ground plane, i.e., a semi-infinite cloud.

t

               'Ibe concentration of radiosenve matenal wnhin this cloud is taken to be uniform and equal to the marirman centerline ground level concentration that would exist in the cloud at the .wwy m dieraner from the point of release.

The whole-body dose is a result of exposure to external gamma radation The whole-body dose due to immersion in a semi-infimte cloud is given by (Reference 8):

                                                 <wk                 ----  <Jon                                            X g = 0.25        x/Q                                                                               [3-1]

s . k .

                                                                                                              ~
                                                                 ~

_., _ a . - ,_

a -a __ System 80+ Deelan Contrtd Documerrt where:

          %,           =      is the whole-body gamma dose from immersion in a semi-infinite cloud for a given time period, rem.

ak 'g= is the activity of isotope i released during a given time period, curies /y g 4 x is the atmospheric dilution factor for a given time period, sec/m' hl

          .Svi
                       =      is the average gamma radiation energy enutted by isotope i per disintegration, MeV/ dis The gamma energies pid in Table 15A-2A include the X-rays and annihilarion gamma rays if they are prominent in the electromagnetic spmu. _                             , j, ,          ,

3.3 Thyroid Inhaladan Dose - -- l

           'Ihe thyroid inhalation dose is obtamed from the following expression (Reference 9):

Dm " X /Q B E Q, - (DCF), s [3-2] )

                                                                                ~
               .- ..              .        . ~ . .          . . . . . .

where:

Dm i thyroid Inhalation dose, rem x/Q = atmosoteric dilution factor for a given time period, ser/MS

                                                --..          a -+ t.             v .. . c. m .u_.-.      .      -.

B ^ =< beesthing rate for a given time period t, m 8/sec ~ ' "" '" Qi = total activity of iodine isotope i released in time period t, curies (DCF), = dose conversion factor for iodine isotope i, rem / curie inhaled l The thyroid inhalanon dose conversion factors and offsite bio.ddug rates used in the above model are given in Table 15A-2B and Table 15A-8, ,@vely. 3.4 Control Roosa Dose

            'Ihe thyroid inhalation, whole-body gamma. and beta skin dose models for the major contribinors to the Control Room dose are described below. The dose to the Control Room c-x --;== due to a pa=adH-
            -N is calculated on the basis of source strength, atmospheric transport, dosimetry and Control Room emergency pio.mu. don and filtration considerations as illustrated in the following equations and used in References 6 and 7.

The thyroid inhalation dose is obtamed from the following expression: D, = E IBR DCF, T'A,(t) di [3-3] ij . r,, W Deegn AdsomW- Aaelsimt Anaheir Aspe f544

i l t i System 80+ Desem conw oocument where: Day = thyroid inhalation dose, rem 4 . d i = isotope index l 3 !, j = time intervalinoex

BR = breathing rate, m 8/sec DCFi = dose conversion factor for lodme isotope i, rem / curie inhaled l _

Aft) = airborne concentration of iodine isotope i at time t (sec), in the Control Room, i curies /rc' , 4 j The thyroid inhalation dose conversion factors and Control Room breathing rate used in the above mode!

;      are ga in Table 15A-2B and Table 15A-8, respectively.

The whol% gamma dose due to inleakage is calcularM using the following q= dan: D, = (R/2) E A, (K,) Ml i wh-: j

j Da = CMf gamma dose, mn J R = radius of an equivalenth ==i=ahaical control room, m i ! l Ai - time i~a- d cancamration of melW i, Ci-br/m' 3-5 I 2 1 (Km). = dose conversion factor for nuclide i, Rem-m /Ci-br definad by q=daa_ 5A-A GC.), = 3.7 x le E. S,C, [3-5] b Se = gamma energy emmad per disintegration for melW i at energy E,3 Mev/ dis 2 Cj = Dux to dose rate conversion factor for gamma energy E3 , rem-cm -sec/Mev-br, Reference 10 3.7x11 = Units cowersion factor, dis-m 2/Ci-sec-cm 2 The gamma energies E, 3 are presented in Tab!: 15A-2A. The beta skin dose due to inleakage is calculated using the following equation (Reference 8): M1 D, = 0.23 E. M [3-6) AppewuW Omegn AssameW AseWow Amafpeer A ge f54-5

                .~. .-                --                -       .    ..        ..       -   - _-- . _ - _              - _ _ _ _ _ - _ _ _

System 80+ Demiser Correof Doewnerrt where:

            'D.                =          beta skin dose from immersion in a semi-infinite cloud, rem
           'Ai                =           time integrated concemration of nuclide i, Ci-sec/m 3 k                                  average beta energy emitted by isotope i, MeV/ dis

? V The 0 ~ include conversion electrons if they are prnminant in the electromagneuc spectrum and / are cw=M as.one-third the ." ' ... ... beta energy for a given spectrum. The values of E, are v.M in Table 15A-2A. 1he whole-body gamma doce to Control Room pe.C due to arcloud external to the Control Room is calailatM using the followmg equation: D, = E (x/Qj E. A, CF) [3-7] s s where: D, = whole-body gamma dose due to external cloud shine, rem x/Q3 - men pharic di& factor for the time periodj, calanlatM at vanous locations at the Control Room (such as the air intakes control room doorway or center of control room roof) assummg a ground level release, sechn8 A, = total activity of avlide i released during time period j, Ci CFi = a dose rate response function for a umt cancentration of arlida i, rem-m2 /Ci-sec 4.0 Computer Codes Used in Accident Analysis lhe computer programs desenbed below have been used to calailata design-basis source terms and radiological cansary==ca of design basis scridante. 4.1 DRAGON (Reference 6) Program DRAGON evaluates the activities, dose rates, and time-WM dose in the reactor building and control room of a arlaar facihty or at an adjacent site following release of halogens and noble gases from some control volume. The fission product release to the at=nsphere, together.with the activities and time-integrated activity concesrations of the halogens which are acesnailatad in the system, are also computed. Site dose calculations performed by DRAGON employ the semi-infinite cloud models suggested by Regulatory Gmde 1.4. The gamma dose in the control room is ce_*M based upon a finite cloud model. Average beta and gamma energies r e used in all dose calculations. Ampoved Des @n A8searW- AeaWmW Ama$udr Auge Fa4-5

System 80+ Design ConeelDocannent ne following describes how these two items are calculated for non-LOCA events (Sections 15.7.3 and 15.7.4 are excluded): e Sensible Heat Calculation: To facilitate calculation of the sensible heat, the following conservative assumptions are made:

               -        'Ihe RCS metal ^wi-me rammina the same as the average coolant w n . during cooldown, e.g., when coolant tw uo     n is reduced by 100*F per hour, the RCS metal m sysa s m s !s FM"cM at the same rate.
                        ^
               -        De RCS heat loss to the containrent will not be ==ned fot- i.e., assummg heat transfer is only in one direction, towards the coolant.
               -        De specific heat of the coolant and of the RCS naal remam constant at their intial higher values thrmghE the cooldown.

/S .

                             -[            -

(T) + g " '587 ter

                                            ~             ~       ~
                                                                         , g .<, d N JcdL,t]             ust I

l 1 g+ g_

                    /g       ,
                                 % (T) =

gad VfD Gao - - + Cg [6-2]

                                    **            P,ws       ,           ?,s4
               "'r.

T =- RCS metal or coolant ^w uc in *F C, = specific heat in Btu /lbm 'F t- = time in hours RV = Reactor Vessel RP = RCS piping PZR = Pressurizer - SG = Steam Generator AnnmentDeatpn neeamma!- Aeawant Anerpeer rene ran-r2

s. -

l Svstem 80+ Denian canearoccament

Table 15A-7 Generated Iodine Spike Appearance Rates

i lastope Cl/See I 131 2.16 I132 3.12 1133 4.58 1 134 5.04 1 135 4.25 1

500 tinur, the egalibrma Appearance Rase 1 Speenlun is used only for.the coutro1 room dose eatenI=ria==. The I-131 'a*ancal apaah equivalent source is used for the site boundary dose enle=8== for those events which do not have fuel damage ~ pedicted. ;

Table 15A-4 Breathing Rates for Adults (Reference 8) i i 11sne frans State of Accident B40 (us'/ase) I see 0 to 8 hrs. 3.47 x 10d i 8 to 24 hrs. 1.75 x 10' i i to 30 days 2.32 x 10d Control 3com_ k JO-idays k 3.47 x lod fb0A1 BC r M 500 times the equihirium Appearance Rate

                %um is used only for the control room dose calculations. The I-131 'echme=1 specification equivalem source is used for the site boundary dose calculations for those events winch do not have fuel damage Protheted.           .

Anprowed Damen noneard- Aeedent AneWu hwel&A-24 (5 2-

i i U st i 9 fll (*} ~ ' . c. ff

                                                                                                  < }/

4 p.v v'E !.

                                      ~]                      CHAPTER 16 4

D LQ g j Ic[ ^ ' * * ,, TECHNICAL SPECIFICATIONS b / O I c v t.< O 5 G PREFACE d Technical Specifications are explicit restrictions on the operation of a commercial nuclear power plant. They are designed to preserve the validity of the plant safety. analysis by ensuring that the plant is operated within the required conditions bounded by the analysis, and by ensuring that equipment assumed to be available for accident mitigation is operable. Technical specifications preserve the primary success path relied upon to detect and respond to accidents. They also complement the concept of defense in depth. Section 182a of the Atomic Energy Act of 1954, as amended (the Act), 47 U.S.C. 2011, at 2232, provides the legislative framework within which

                 -technical specifications are required. Section 182a of the Act requires in part:
                           "In connection with applications for licenses to operate production or utilization facilities, the applicant shall state such technical specifications, including information on the amount, kind, and source of special nuclear material required, the place of use, the specific characteristics of the facility,-

and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization or production of special nuclear material will... provide adequate protection to the health and safety of the public. any license issued.."Such technical specifications shall be a part of Th: regulatory framework implementing Section 182a of the Act is the NRC's re; lation 10 CFR 50.36, " Technical Specifications." pr: ides in part that each operating license: This regulation

                           ...will include technical specifications ...(to) be derived from the analysis and evaluation included in the safety analysis report, and amendments thereto...and may also include such additional technical specifications as the Commission finds appropriate."

This set of System NUREG-1432, Standard 80+ Standard Technical Specifications is based on Technical Specifications, Combustion Engineering Plants, and establishes these conditions and limitations for the System 80+ Standard Design. This set of technical specifications is intended to be used as a guide in the NRC's development of the plant-specific technical specifications issued with the operating license. The values provided in brackets [ ] are preliminary.

ec1itsnamma LIST OF TABLES CHAPTER 16 Table Sublect

        <     1.1-1           MODES 3.1.12 1
       '                      REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT OPERATIONAL MODES FOR K, > 0.98 3.1.12-2
       '                      REQUIRED MONITORING FREQUENCIES FOR EACKUP BORON DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT OPERATIONAL MODES FOR 0.98 2 K, > 0.97
     ,       3.1.12-3 REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT OPERATIONAL MODES FOR 0.97 2 K, > 0.%

3.1.12-4

   '                         REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT OPERATIONAL MODES FOR 0.% 2 K, > 0.95 3.1.12-5 REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT OPERATIONAL MODES FOR K, s 0.95
      ,     3.3.1-1 REACTOR PROTECTIVE SYSTEM INSTRUMENTATION - OPERATING
      ~    3.3.2 1 REACTOR PROTECTION INSTRUMENTATION - SHUTDOWN
     ,     3.3.5-1 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM                I INSTRUMENTATION 3.3.6-1 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM LOGIC AND MANUAL TRIP
    -      3.3.8-1 ALTERNATE PROTECTION SYSTEM
   -      3.3.10-1 CONTAINMENT BYPASS INSTRUMENTATION (STEAM GENERATOR TUBE RUPTURE) 3.3.11-1 POST ACCIDENT MONITORING INSTRUMENTATION
    -     3.3.12-1 REMOTE SHUTDOWN INSTRUMENTATION AND CONTROLS
         -Mr+ .3 [ ./O - / REQUIRED MINIMUM SET OF HMS IGNITERS 3.7.1-1

[ VARIABLE OVERPOWER TRIP] SETPOINT VERSUS OPERABLE MAIN STEAM SAFETY VALVES P

CESGAWMuifegi. 1 LIST OF TABLES (Continued) CHAPTER 16 Table Subiect f

 /   3.7.1 2   MAIN STEAM SAFETY VALVE LIFT SETTINGS 3.8.1 1   DIESEL GENERATOR TEST SCHEDULE 3.8.6-1   BATTERY ELECTROLYTE REQUIREMENTS         .

5.2.2 1 MINIMUM SHIFT CREW COMPOSITIO b [ Single Unit Facilg]

                                         '\ i ')

l l vii e ent U 12/31/93

l DEhMTnPM&v l LIST OF FIGURES CHAPTER 16 Figure Subiect 3.1.5-1 REQUIRED POWER REDUCTION AFTER CEA DEVIATION 3.4.3-1 A RCS PRESSURE AND TEMPERATURE LIMITS (HEAT-UP) 3.4.3.1B RCS PRESSURE AND TEMPERATURE LIMITS (COOLDOWN)

.5.4 3 1 ALLOWED IRWST TEMPERATURE VS. CONTAINMENT ATMOSPHERE TEMPERATURE 3.7.20-1 DISCHARGE BURNUP VS. INITIAL ENRICHMENT FOR REGION II RACKS 4.1 1 SITE AND EXCLUSION AREA BOUNDARIES 4.1-2 LOW POPULATION ZONE 3

           ,,3_,   g,. a,. & Go l" Y DOSE Ejilt)IV}} LENT .T- o 3 l cpaa f,c AcH47 L.% .1 L% .cas % a ,-f R A 7 e a TH ef? MAL Po wG g             .f y couland Spec,-fic Ac+;v,*4,, > l. O m;a eo ls'  m Oos e EQvwAL ENT f- I S } ,

viii 3

System 80+ ' Deslan ControlDocument 2 H TABLE OF CONTENTS j L 1.0 USE AND APPLICATIONS ..................... 1.1-1 ? 1.1 Definitions * ,

                                                      .......................                                           1.1-1 1.2                                                                                                                               k Logical Connectors . . . . . . . . . . . . . . . . . . . .                        1.2-1 1.3                                  Completion Times . . . . . . . . . . . . . . . . . . . . .                         1.3-1                              -

1.4 Frequency ........................ 1.4-1 2.0 SAFETY LIMITS (SLs) ..................... 2.0-1 2.1 Safety Limits ...................... 2.0-1 2.2 Safety Limit Violations ................. 2.0-2 - -- 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY . . . . . 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.0-4

                                                                                             .........                         o           n,
                                                                                                                               *~            .

3.1 3.1.1 3.1.2 REACTIVITY CONTROL SYSTEMS . . . . . . . . . . . . . . . . . . SHUTDOWN MARGIN 'SDM) .................. 3.1-1 3.1-1 ( .

                                                                                                                                             ?
                                                                                                                                             ')

Reactor Trip Circuit Breakers (RTCB) . . . . . . . . . . . 3.1-2 u 3Ih 3.1.3 Reactivity Balanct .................... 3.1-3 a f 5.1.4 l.1.5 Moderator Temperatere Coefficient (MTC) ......... Control Element As.'embly (CEA) Alignment . . . . . . . . . 3.1-5 4;%4+ t 3.1-7 , *

r 1.1.6 Shutdown Control F,lement Assembly (CEA) Insertion Limib . 3.1-12 s +.i -

l.l.7 1.1.8 Regulating Control Element Assembly (CEA) Insertion Limits 3.1-14 Part Strength C'>ntrol Element Assembly (CEA) Insertion 1[ 4- "# Limits . . . . . . . . . . . . . . . . . . . . . . . . 3.1-18 l.1.9 Special Test dxception (STE)-SHUTDOWN MARGIN (SDM) . . . 3.1-20 n.1.10 Special Test Exceptions (STE)-MODES 1 and 2 . . . . . . 3.1-22 l.l.11 Special Test Exception - CEDMS Testing . . . . . . . . . 3.1-24 1.1.12 Boron Dilution Alarms . . . . . . . . . . . . . . . . . 3.1-26 ; l.2 POWER DISTRIBUTION LIMITS l.2.1

                                                                        ................                        . 3.2-1 i.2.2 Linear Heat Rate (LHR) . . . . . . . . . . . . . . . .                      . 3.2-1 Planar Radial Peaking Factors (Fxy)                 ...........                   3.2-3 l.2.3                                AZIMUTHAL POWER TILT (T,)
.2.4                                                                           ................                       3.2-5 Departure From Nucleate Boiling Ratio (DNBR) . . . . . . . 3.2-8
.2.5                                AXIAL SHAPE INDEX (ASI) . . . . . . . . . . . . . . . .                          3.2-10
.3                             INSTRUMENTATION
.3.1                                                 .......................                                          3.3-1 Reactor Protective System (RPS) Instrumentation - Operating 3.3-1
.3.2                                Reactor Protective System (RPS) Instrumentation - Shutdown 3.3-13

.3.3 Control Element Assembly Calculators (CEACs) . . . . . .

.3.4                                                                                                                 3.3-20 Reactor Protective System (RPS) Logic and Trip Initiation 3.3-23
.3.5                               Engineered Safety Features Actuation System (ESFAS)

Instrumentation . . . . . . . . . . . . . . . . . . . . 3.3-26 j .3.6 Engineered Safety Features Actuation System (ESFAS) Logic and Manual Initiation . . . . . . . . . . . . . . 3.3-32 .3.7 Diesel Generator (DG) - Loss of Voltage Start (LOVS) . . ! .3.8 3.3-37 Alternate Protection System (APS) . . . . . . . . . . . 3.3-40 (continued) sw ro > mes w.r *s,.e p.,. i _ _ _ . . _ _ _ _ _ . . - , . , - - - . m--

Syntem 80+ Design ControlDocument Definitions 1.1 1.1 Definitions (continued) SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All full length Control Element Assemblies (CEAs) (shutdown and regulating) are fully inserted except for the single CEA of highest _

reactivity worth which is assumed to be fully withdrawn. However, with all CEAs verified fully inserted by two independent means, it is not necessary to account for a stuck CEA in the SDM calculation.

b. In MODES 1 and 2, the fuel and moderator temperature are changed to the [ nominal zero power design level].

With any CEAs not capable of being fully inserted, the reactivity worth of these CEAs must be accounted for in the determination of SDM. STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the Q y,L.. y testing of one of the systems, subsystems, t 4 , a. channels, or other designated components during a the interval specified by the Surveillance Frequency, so that all systems, subsystems,

i. h % 4 w 3 49>

CHANNELS, or other designated components are v .2 s. -I ( tested during a Surveillance Freowncy intervals, 4 u4 where a is the total number of s sms, j 4 subsystems, components, or other - ignated g M $ {} components in the associated funt _ 1. E g jgh THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant. iTRIP CHANNEL A TRIP CHANNEL is defined as a set of interconnected hardware and software components that process an identifiable sensor signal to produce the trip or alarm signal associated with the sensor. This includes the sensor, data acquisition, signal conditioning, data transmission, software and all transn!i ssion lines (continued) Approved Desigre Meteria!- 16.1 Tech Spec Page 1,1-8

System 80+ Deskn ControlDocument

Logical Connectors 1.2 1.2 Logical Connectors

                                                                                                                        .l EXAMPLES         EXAMPLE 1.2-1 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Verify -- l A.2 Restore In this example the logical connector M is used to _ indicate that when the Condition A, both required Action A.1 and A.2 must be completed. 1 I EXAMPLE 1.2-2 , ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LC0 not met. A.1 Trip... E A.2.1 Veri fy. . . l M A.2.2.1 Reduce... E (continued) dl mb; $ , C N L or % *

                                                                                               % y (r
                                                                                                   ~s r                                                                                                   a ! i(' L:

w o + u.m re., r.e, sn r.,. ,.2 2 dC ' t

I 3 Sv-t m 80 + oestan contros Document Completion Times I.3

1.3 Completion Times DESCRIPTION EXAMPLE 1.3-5 (continued) t The Note allows Condition A to be entered separately for 4 each inoperable valve, and Completion Times tracked on a per valve basis. When a valve is declared inoperable, Condition l 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 and are __ j tracked for each valve. . 2 l If the Completion Time associated with a valve in Condition 4 A expires, Condition B is entered for that valve. If the l Completion Times associated with subsequent valves in i Condition A expire, Condition B is entered separately for J each valve and separate Completion Times start and are i tracked for each valve. If a valve that caused entry into ; j Condition B is restored to OPERABLE status, Condition B is ; exited for that valve. i Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply. : 1 3 EXAMPLE 1.3-6 , i ! 4 ACTIONS ! CONDITION REQUIRED ACTION COMPLETION TIME 9 A. One channel A.I Perform SR Once per 8 hours

inoperable 3.x.x.x.

1

)                                           98                                                                    l J

T fhours b l A.2 Reduce THERMAL POWER to s 50% h l j RTP. i

(continued)

} Novowed Desfon Ateneniet- 16.1 Tech Soc pay ,y,3.g i !

'ystem 80+

Deslyn ControlDocument Completion Times ,

1.3 I .3 Completion Times (continued) IECIAL EXAMPLE 1.3-7 l JMPLETION l 8MES ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Verify affected I hour subsystem subsystem . .. inoperable. isolated. Once per 8 hours thereafter DB A.2 Restore subsystem 72 hours to OPERABLE status. r-B. Required B.1 Be in MODE 3. 6 hours Action and associated M Completion Time not B.2 Be in MODE 5. 36 hours met. Required Action A.1 has two Completion Times. The 1 hour Completion Time begins at the time the Condition is entered and each "Once per 8 hours thereafter" interval begins upon performance of Required Action A.I. If after Condition A is entered, Required Action A.1 is not met within either the initial I hour, or any subsequent 8 hour interval, from the previous performance (including the 25% extension allowed by SR 3.0.2), Condition B is entered. The completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, A. v provided the Completion Time for Required Action A.2 has not 9 .- expired. 4 1(. $ 4 g 4' (k g.

                                                                                              <      n N
                                                                                                  .f continued) +
    ,          . . , r.a sn .
                                                                             ,.s. , ,
                                                                                        ,5        QQ

System 80+ Deslan ControlDocument Completion Times l 1.3 j 1.3 Completion Times I * ' i SPECIAL EXAMPLE 1,3.-Z

                                       .    (continued)

COMPLETION TIMES Since the second Completion Time of Required Action A.1 has a modified " time zero" (i.e., after the initial I hour, not from time of Condition entry), the allowance for a ; Completion Time extension does not apply. IMMEDIATE I - COMPLETION TIME pWhen "Immediately" is used as a Completion Time,jg,,7 the Required Action should be pursued without delay and in a , controlled manner. On li n t M? C v #-- i I Aweved Design Metenfel 16.1 Tech Spec 4 y,3,gg

        .                                                  _               =                               -

System 80+ Design contror Document SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS y 3.1.1 SHUTDOWN MARGIN (SDM) h0I -) epoJ'1 LC0 3.1.1 a. SDM shall be 2 [6.5]% Ak/k; and either A OL b.1 With reactor trip circuit breaker (RTCBs) closed: the estimated critical position shall be within the limits 4/) i of LCOs 3.1.6 (" Shutdown Control Element Assembly (CEA) he g6 j Insertion L1mits") and 3.1.7 (" Regulating Control Element Assembly (CEA) Insertion Limits"); or h@gelghg _ b.2 With RTCBs open: K,.3 shall be < 0.99. APPLICABILITY: MODES 3, 4 and 5. l ACTIONS . CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1 Initiate boration to 15 minutes restore SDM to within limit. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM is 2 [6.5]% Ak/k. 24 hours i Anwered Denka Atatorial- 16.3 Tech Spec pag 2. p.y

Sv-tem 80+ oestan conoot Document T 3.2.3 , gyEILLANCEREQUIREMENTS SURVEILLANCE FREQUENCY - SR 3.2.3.1 -------------------NOTE-------------------- Only applicable when COLSS is out of service. With COLSS in service, this _ parameter is continuously monitored. 12 hours ate T, and verify it is within the g I'AD SR 3.2.3.2 Verify COLSS azimuthal tilt alarm is 31 days actuated at a Tq value less than the T, value used in the CPCs. SR 3.2.3.3 Independently confirm the validity of the 31 days

               .COLSS calcu' lated T, by use of the incore detectors.                                                                                                 .

l l wndonw unww 1s.3 run sm m s.2 7 l

Syst m 80+ Deslan Contr*lD cument - DNBR 3.2.4 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME B. DNBR outside the B.1 Restore DNBR to region of acceptable 4 hour {3 4(' within limit. ( operation when COLSS- \ is out of service. C. Required Action and C.1 Reduce THERMAL POWER 6 hours associated Completion to s 20% RTP. Time not met. SURVEILLANCE REQUIREMENTS - SURVEILLANCE FREQUENCY SR 3.2.4.1 --------------------NOTE------------------- Only applicable when COLSS is out of service. With COLSS in service, this parameter is continuously monitored. Verify DNBR, as indicated on all OPERABLE 2 hours i DNBR CHANNELS, is within the limit of Figure 3.2.4-1 or 3.2.4-2 of the COLR, as

applicable.

i l SR 3.2.4.2 Verify COLSS margin alarm actuates at a 31 days l 3 THERMAL POWER level equal to or less than i the core power operating limit based on DNBR. ' wmDeew neeww 1s.2 recs spec r ,. 3.2.s i

                                                                     -                             __J

Syntem 80+ Deslan Control Document RPS Instrumentation - Operating 3.3.1 1 ACTIONS (continued) I CONDITION REQUIRED ACTION COMPLETION-TIME l C. One or more' Functions C.1 ---------NOTE-------- with one operating Applies only to ; bypass removal Functions 2, 4, 13, I function inoperable, and 14 in Table

_i- '

3.3.1-1

4 (

j -1 hour :d 1 Verify operating i3 ' bypass removal M : f2 ; function is not in g effect. Once per 12 hours thereafter . e f :. E2 s p. , gg A - O ;

                                                                          -1 wour                    W ,5 0          ,

C.2.1 Place affected - I l automatic TRIP CHANNEL in bypass or trip. M '6 Prior to <m C.2.2 Restore operating entering MODE 2 ! bypass removal function and following next MODE 5 entry r

                                                                                                            ,,  }3[g i

l associated automatic p' , TRIP CHANNEL to 4 . OPERABLE status. W '". t D. One or more Functions D.1 --------NOTES-------- with two operating 1. LCO 3.0.4 is not bypass removal applicable, functions inoperable.

2. Applies only to Functions 2, 4, 13, and 14 in Table 3.3.1-1.

_____________________ j l (continued) l I Approwd Desie Manwier- ts.3 Tech spec re y 2.2 2

Syst m 80+ Deslan Control Document RPS Instrumentation - Operating 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. (continued) D.1 (continued)- Verify operating I hour - 4 bypass removal functions are not in y 5* AND effect. 4 r Once per 12 - Qg }

                                                                                   ._       hours thereafter             h4f L     $

D.2 Place one affected I hour q W <s automatic TRIP CHANNEL in bypass and [@ place the other in trip. , E. Required Action and E.1 Be in MODE 3. 6 hours associated Completion Time not met. j 4 . Anwowmf Design nietwW 16.3 Tech Spec pp 3,3,3

Syntem 80-t- Design ContclDocument RPS Instrumentation - Operating 3.3.1 SURVEILLANCE REQUIREMENTS

.---------------------------------NOTE------=- - - - - = - -

~ '

Refer to Table 3.3.1-1 to determine which SR shall be performed for each RPS " ' Function.

_- __- --- _ -_ _ -----===_

SURVEILLANCE FREQUENCY

12 hours SR 3.3.1.1 Perform CHANNEL CHECK. SR 3.3.1.2 ---------------------NOTE------------------ Not required to be performed until 12 hours after THERMAL POWER 2 70% RTP. 12 hours Verify total Reactor Coolant System (RCS) flow rate as indicated by each CPC is less than or equal to the RCS total flow rate. D e 40 Wg SR 3.3.1.3 Check the CPC autorestart count is less 12 hours than three. N t N SR 3.3.1.4 Verify PPS cabinet temperatures are below 12 hours the high limit. (continued) l knproved Denkm Materiel 16.3 Teeh Spec Page 3.34

System 80+ Design C*ntrol Document RPS Instrumentation - Operating 3.3.1 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY

                                                                                                                  % _4. ,o         .

SR 3.3.1.5 -- 1.

                                 - ==--
                                           -------NOTES---------------------

Not required to be performed un+,il 12 34 hours after THERMAL POWER 2 20% RTP. 1

2. The daily calibration may be suspended during PHYSICS TESTS, provided the calibra+. ion is performed upon reaching ..

each major test power plateau and prior to proceeding to the next major test power plateau.

                                =_________________________________-_--__

Perform heat balance calibration and adjust 24 hours the linear power level signals and the CPC addressable constant multipliers to make the CPC AT at power and CPC nuclear power calculations agree with the calorimetric, if the absolute difference is 2 [2] %. Sen adjust excore nuclear power to agree with CPC nuclear power if the absolute l difference is 2 [2%). I SR 3.3.1.6 = = == = ------ --NOTE--------------------

w Not required to be performed until 12 hours b after THERMAL POWER 2 70% RTP.

1 s p -31 days .t.

                                                                                                                  #    =D Verify total RCS flow rate indicated by                                             g       i y

each CPC is less than or equal to the RCS flow determined by calorimetric 'Q% l calculations. S e. t 3 t l (continued) I l l Apperved Design Material 16.3 Tech Spec page 3.3 5

System 80+ Deslan ControlDocument RPS Instrumentation - Operating i 3.3.1 ; SURVEILLANCE REQUIRENENTS (continued) ! f SURVEILLANCE FREQUENCY ;

                                                                                                                                          -l SR 3.3.1.7        ---------------------NOTE-------------------                     731 days                                               l Not required to be performed until 12 hours                       I l

l after THERMAL POWER 2 15% RTP. I y g,

Verify linear power gains of the excore - ~t < N ! l detectors are consistent with the values Q i i used to establish the shape annealing- l matrix elements in the CPCs.- l ' SR 3.3.1.8 ---------------------NOTES------------------

1. The CPC CHANNEL FUNCTIONAL TECT shall i l-include verification that the correct-I values of addressable constants are installed in each OPERABLE CPC.

2. Not required to be performed for Logarithmic Power Level-High until 2 hours after reducing THERMAL POWER !

below [1E-4%) RTP and only if r_eactor trip circuit breakers (RTCBs) are closed. k Perform CHANNEL FUNCTIONAL TEST. 92 days i SR 3.3.1.9 ---------------------NOTE------------------- Neutron detectors are excluded from CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION of the power 92 days range neutron flux channel. (continued) named neew neeuw- se.s recs spec re ,e s.s.s [. _

?vstem 80+ Desinn ControlDocument RPS Instrumentation - Operating 3.3.1 IRVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY ad

                                                                                               + --    o%          *l 7
                                                                                                    ~7             -

5R 3.3.1.10 S L Perform CHANNEL CALIBRATION on A/D Reference Sources. [18] months b a i 2R 3.3.1.11

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

Neutron detectors are excluded from CHANNEL - - CALIBRATION. Perform CHANNEL CALIBRATION on each TRIP [18 r.onths] l CHANNEL, including operating bypass removal l functions. 36 3.3.1.12 Perform a CHANNEL FUNCTIONAL TEST on each [18 months] CPC CHANNEL. I

IR 3.3.1.13 Using the incore detectors, determine the Once after each l ! shape annealing matrix elements to be used refueling prior by the CPCs, to exceeding 70% RTP

!R 3.3.1.14  Perform CHANNEL FUNCTIONAL TEST on each                    Once within

automatic operating bypass removal 92 days prior function. to each reactor startup

' 4 ! l' (continued) M Du> MeswW 16.3 Tuh Spu Page 3.3 7

s-System 80+ Deslan ControlDocument RPS Instrumentation - Operating 3.3.1 ! SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY l v4 1

                                                                                          *3        i SR 3.3.1.15   ---------------------NOTE------------------                                   4 Neutron detectors are excluded.                                             "

___________________________________________ N Verify RPS RESPONSE TIME is within limits. [18] months on a STAGGERED TEST BASIS -- -- l 1 I I 1 I I i ( l I l 1 i monw meraw ts.3 r.as se.c r. ,. 3.s.a

l System 80+ l ' Deslan controlDocument RPS Instrumentation - Operating 3.3.1 Table 3.3.1-1 (Page 1 of 4) Reactor Protective System Instrunentation Operating i APPLICABLE MODES OR OTHER SPECIFIED sVRVE!LLANCE FUNCTION CONDITION REQUIREMENTS ALLOWABLE VALUE , i

1. Variable overpower - Nigh 1,2 SR 3.3.1.1 s (112.73% RTP st 3.3.1.4 SR 3.3.1.5 st 3.3.1.7 SR 3.3.1.8 i SR 3.3.1.9 l

[sR 3.3.1.103 ' sR 3.3.1.11 i sa 3.3.1.15

2. Logarithmic Power Level - High(a) 2 SR 3.3.1.1 s [0.0181% RTP st 3.3.1.4 SR 3.3.1.8 Ist 3.3.1.103 l SR 3.3.1.11 '

SR 3.3.1.14 ! SR 3.3.1.15

3. Pressurizer Pressure - High 1,2 sa 1.3.1.1 5 I2370 pois) l SR 3.3.1.4 '

st 3.3.1.8 Ist 3.3.1.101 st 3.3.1.11 st 3.3.1.15

i. Pressurizer Pressure - Low Trp- "

1,2 SR 3.3.1.1 Trip t (1825 psla); SR 3.3.1.4 Trip operating b.Y

                                                                                                                      S'l F ^ * ,,

SR 3.3.1.8 Bypass RemoveL t Ist 3.3.1.101 (500 psia); [ {pon - - - SR 3.3.1.11 step 5 (400 petal, st 3.3.1.14 Floor n [300 pala) g , , , ,. , i st 3.3.1.15 ggg . , , l l

5. Contairinent Pressure - High

1,2 SR 3.3.1.1 st 3.3.1.4 s [2.7) psig ' lia e vf* 4 5 sa 3.3.1.8 gA, w y [st 3.3.1.10) st 3.3.1.11 _ st 3.3.1.15 (continued) (a) Trip may be bypassed when THERMAL POWER is > (1E 43% RTP. Operating bypass shall be automatically removed when THERMAL POWER is s [1E-43% RTP. Trip may be manually bypassed during physics testing pursuant to LCO 3.4.16, "RCS Loops - Test Exception." Approved Design nieteriel.16.3 Tech Spec Page 3.3-9

_ . . . ~ . _ - _ _ _ _ _ .. - - - . _ - - . _ _ ___ _ . _ _ _ . . . _ _ _ . -_ System 80+ & ControlDocument RPS Instrumentation - Operating 3.3.1 Table 3.3.1 1 (Page 3 of 4) Reactor Protective system Instrumentation - Operating APPLICA8LE MODES OR -

OTHER SPECIFIED SURVEILLANCE FUNCTION CONDIT10W REQUIREMENTS ALLOW 48LE VALUE

11. steen Generator 82 Level-High 1,2 st 3.3.1.1 s (90.8% Narrow l SR 3.3.1.4 Rensel t

st 3.3.1.8 tsa 3.3.1.10) % SR 3.3.1.11 g sa 3.3.1.15 C m

                                                                                                                                                     +       O-

12. Reactor Coolant Flow - Lew(d) #

                                                                            .1,2 V uQ SR 3.3.1.1      Rates s (*)

l SR 3.3.1.4 psi /sec.; , l j SR 3.3.1.8 tsa 3.3.1.10) Floorst (*) pold; step: [*J pel py ,, a L sa 3.3.1.11 d st 3.3.1.15 -* dg el e as E-kh l 13. Local Power Density - Nigh(') 1,2 st 3.3.1.1 s (21.0) kW/ft > st 3.3.1.2 l st 3.3.1.3 st 3.3.1.4 st 3.3.1.5 st 3.3.1.6 st 3.3.1.7 SR 3.3.1.8 l SR 3.3.1.9 [st 3.3.1.103 SR 3.3.1.11 ( st 3.3.1.12 l SR 3.3.1.13' st 3.3.1.14 st 3.3.1.15 (continued) V:Lue to be determined by system detait design. (d) The Reector Coolant Flow Low trip setpoint varies with reactor power and is rate limited with a preset low power level. The Reector Cootent Flow-Low trip setpoints may be manuelty adjusted een THERMAL POWER is < [1E-43% RTP. (s) Trip may be bypassed when THERMAL POWER is < tie 4]E RTP. Operating bypass shall be automatically removed when THERMAL POWER is t (1E-41% RTP. During testing pursuant to LCO 3.4.16, trip may be byps; sed below [5%) RTP. Operating bypass shall be automaticatty removed when THERMAL POWER is > (5%) RTP. I 1 l W M M W N W

16.3 IeCh W toppe 3.311

?

i System 80+ N '-- concer Docenent ' RPS Instrumentation - Shutdown 3.3.2 ' ACTIONS (continued) l CONDITION REQUIRED ACTION COMPLETION TIME : j B. One or more functions B.1 --------NOTE--------- ' with two TRIP CHANNELS LCO 3.0.4 is not inoperable. applicable.

                                                                       ' Place one TRIP                       1 hour                                                  ,

CHANNEL in bypass and _ i place the other in trip. C. One or more functions C.1 -------NOTE----------- I with one automatic Applies only to *1 8} r, ; cperating bypass Functions 1, 5, and 6 c. ; 4 : removal function in Table 3.3.2-1. m + ,~ t inoperable. --------------------- j ,M 4 ,8 p 1 hour (

                                                                                                                                       )
                                                                                                                                                     ~3         e Verify operating                                                       '-            t bypass is not in                     8llQ                      l:3            !

2 effect. *

                                                                                               ,             Once per 12               S 4=$Q;'*. g.        .        l QB                                      ~\            hours thereafter          g       ;n.

l

                                                                                                           -1 hour w ,J TPvy =0i l

C.2.1 Place affected -I l automatic. TRIP CHANNEL in bypass or trip. 8!10 p Prior.to C.2.2 Restore operating entering MODE 2 bypass removal following next function and MODE 5 entry associated automatic TRIP CHANNEL to OPERABLE status. (continued) Annvemtan6n noenw w.2 rech spec w s.s.u

l l ! System 80+ Deslan ControlDocument i RPS Instrumentation - Shutdown 3.3.2 FACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME l-

D.

One or more Functions D.1 --------NOTES-------- with two automatic 1. LCO 3.0.4 is not i operating bypass applicable. rg , 2 r:noval function s . ' j inoperable. 2. Applies only to 1 + 4 % Functions 1, 5 3 g1 ~ and 6 in Table $

                                                                                                                                         'a     l v                 9 i                                                      3.3.2-1.
                                                      ==- -------------
                                                                                                                 ?gy y

3 yj Verify operating I hour # 5., n bypass is not in 4 ; ?. d effect. AND i. g

                                                                                                                          .2 i      '   l QB                                \           Once per 12
                                                                               \          hours thereafter D.2 Place one affected                        I hour automatic TRIP CHANNEL in bypass and                                                                           .

place the other in trip. 1 1 E. R quired Action and E.1 Open all RTCBs. I hour l associated Completion l Time not met. , I 1 enuwowenf Deefon Ateneriel- 16.3 Tech Spec 9 3,3.gy

System 80+ Design CrntrolDocument RPS Instrumentation - Shutdown 3.3.2 SURVEILLANCE REQUIREMENTS (continued) SURVEILLANCE FREQUENCY SR 3.3.2.5 Perform CHANNEL FUNCTIONAL TEST on each 92 days automatic operating bypass removal function. SR 3.3.2.6 Perform CHANNEL CALIBRATION on A/D [18] months __ ___ Reference Sources. ___ SR 3.3.2.7 --

                                                             =-----------NOTE--------------------

Neutron detectors are excluded from CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION on each TRIP (18] months CHANNEL including operating bypass removal l function. SR 3.3.2.8 --------------------NOTE-------------------- Neutron detectors are excluded. Verify RPS RESPONSE TIME is within limits. [18] months on a STAGGERED - TEST BASIS y b u T A 'E, e 1 N svs is vs 0.

                                                                                                                                                                      'O    %     3       j
                                                                                                                                                                     'f}~t\0 I

kywond Design Matedal.16.3 Tech Spec Page 3.3-17

l Deslan control Document l , System 80+ L J RPS Logic and Trip Inititation ,

                                                                                                                                            "   ~

3.3.4 s ACTIONS (continued)

                                                                                   / {COMPLETIONTIME CONDITION                                  REQUIRED ACTION Y

Open all RTCBs. {/ 48 hours 4y ,

        ---------NOTE---------                   B.1                                                                                     g           l B.

RTCBs associated with - t one inoperable CHANNEL ' l I may be closed for up ' to I hour for the l performance of an RPS CHANNEL FUNCTIONAL ; TEST.

          ......................                                                                                                              9 One CHANNEL of RTCBs, one MANUAL TRIP                                                                    g CHANNEL, or two RPS LOGIC CHANNEL                                                               ,/                                                            l inoperable in MODE 3,                                                    Y 4, or 5.                                                                                                                                 l V                                                               l Open the affected             Immediately C. Two CHANHELS of RTCBs,               C.1 two MANUAL TRIP                                RTCBs.

l CHANNELS. or two RPS i 10 Git CHANNEL affecting the same 191P (10 gneperable. 1 l ] 6 hours

                 %. ' ' =t k t t oe aM               D.1      Se in MODE 3.

a n y ' .. .t t ams t e s s oa

.. .* w t o w a e, Ng
- =* mt i t, 2 opea all RIC8s. 6 hours

                    '.R
                                                    )
;                     m.        . . . . . . . .

3

                           -.     . ..          w.

se es aut m st

6 w 44

                                        - . . . s-3

( :- 1 % .e k e

System 80+ Deslan ControlDocument ESFAS Instrumentation l 3.3.5 l- ACTIONS (continued) i i l CONDITION REQUIRED ACTION COMPLETION TIME i C. One or more Functions C.1 Verify operating I hour 1 with one automatic bypass function is ; operating bypass not in effect. E i removal function . inoperable. gg Once per 12 hours thereafter .. C.2.1 Place affected I hour automatic TRIP CHANNEL in bypass or trip. 1 E C.2.2 Restore operating Prior to , bypass removal entering MODE 2 function and following next associated automatic MODE 5 entry. TRIP CHANNEL to , OPERABLE status. '

D. One or more Functions D.1 --------NOTE---------- with two automatic LCO 3.0.4 is not ' ! operating bypass applicable. ! removal function --- --------------- ' inoperable. Verify operating bypass functions are I hour 't not in effect. 4 M j e4 L QB \ Once per 12 f 1 %j 1 hours thereafter ( 4( gh D.2 Place one affected I hour [ 2U automatic TRIP CHANNEL in bypass and Q, *3 J** .c place the other in * # trip. (continued) Anwmd on> unaww- 1e.2 recs spec r ,, s.a.n l

t

l System 80+ Design Control Document ESFAS Instrumentation 3.3.5 , 4 SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY SR 3.3.5.2 Perform CHANNEL FUNCTIONAL TEST. 92 days 4- 3 SR 3.3.5.3 Perform CHANNEL CALIBRATION [18] months 1s G T ' I SR 3.3.5.4 Verify ESFAS RESPONSE TIME is within [18] months on ' limits. a STAGGERED TEST BASIS SR 3.3.5.5 Perform CHANNEL FUNCTIONAL TEST on each Once within 92 automatic operating bypass removal days prior to - function. each reactor-startup i 1 i l l w o

m.,w. te., r. ,,,,,.,.,,

                  . .- . ~ . - -                 - _ -        _ - -              _. _   _     .- -           _.                        -- --                   --,

I I i System 80+ Deslan Control Document $ ESFAS Instrumentation 3.3.5 1

Table 3.3.5-1 (Page 1 of 2)

' ENGlhEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION s APPLICABLE MODES OR . . FUNCTION OTHER SPECIFIED ALLOWASLE VALUE CONDITIONS

1. Safsty injection Actuation Signal 7 c8--a "

e

4. Cont > nsent Pressure - Nis 1,2,3,4 s (2.73 pois

z'<*P Trip t (1825 psia);

b. Pressuriter Pressure - Low *) 1,2,3,4 Trip Operating sypass Removet t (500 pale); Step 5 (400 pele;; Floor E [300 pale) 4 (D.

3 t)

                                                                                                                                                        *fg L
                                                                                                                                           .T     ,
                                                                                                                                                      . ILt

2. Contsirument Spray Actuation ** h 8I'"'I

e. Containment Pressure - High 1,2,3,4 5 18.5 psig)

Nh k Migh l 1 l

3. Containment Isolation Actuation Signet

a. Contalrunent Pressure - Nigge) 1,2,3,4 s (2.7) psig

b. Pressuriser Pressure - Low 1,2,3,4 Trip t (1825 psia);

Trip operating sypass Removal n (500 ( pale); Step 5 (400 pala); Floor t (300 b 3 psia) U t

4. Main Steam Isolation Signet y

a. Ste Generator Pressure - 1, 2(c), 3(c), 4(c) g gg43 p g,j Low )

b. Contefrument Pressure

Nish 1, 2(C) () s (2.7) psis

c. Steam Generator Level High 1, 2 IC), 3(c), 3(C), 4 c,4(C)s (90.8% Narrow Range)

(continued) (s) Thz setpoint any be decreased to a miniman value of [300) psie, es pressuriser pressure is reduced, provided the mergin between pressuriser pressure and the setpoint is maintained 5 (400) psia. Trips may be bypassed when pressuriser pressure is < (400) psia. sypass shall be automatically removed when pressurizer pressure is t (500) psia. The setpoint shotL be automatically increased to the normet setpoint es pressuriser pressure is increased. (b) The s2tpoint may be decreased as steam pressure is reduced, provided the margin between steam pressure and the setpoint is maintained 5 (200) psig. The setpoint shall be automaticatty increased to the normet sitpoint as steam pressure is increased. (c) Th3 Main Staem Isolation Signet (MSIS) Function (Steam Generator Pressure Low, Contaltunent Pressure - High, and Stcan Generator Level High) signals is not required to be OPERAsLE when all associated valves isstated by 13e MSIS Function are closed and [ deactivated). Amareved Dee> MeterW 1L3 Tech Spec pope 3.3 30

r Sv~ tem 80+ Design ControlDocument ESFAS Instrumentation 3.3.5 Table 3.3. Sal (Page 2 of 2) ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION APPLICAgLE MODES OR , , FUNCTION III S

5. Emergency Feeduster Actuation ,h Signal SG #1 (EFAS-1) s t Ch

e. Steen Generator Level - Low 1,2,3 t (23.4% Wide Range) \'

b. Steam Generator Level - High 1,2,3 s [53.4% Narrow Range) h g h

6. Emergency Feochseter Actuation Signal SG #2 (EFAS 2) * ,5 4\ V/

c. Steen Generator Level - Low 1,2,3 t (23.4% Wide Range) N

b. Steam Generator Level - Migh 1,2,3 5 (53.4% Narrow Range]

AnerovedDeelsn Meterial 16.3 Tech Spec Page 3.3 31 w.

Syntem 80+ o+n controlDocument DG - LOVS 3.3.7 ACTIONS .' M bia;d)- CONDITION REQUIRED ACTION COMPLETION TIME B. (continued) B.2 Restore two or 8 hours more CHANNELS to OPERABLE status for each Function.

C. One or more Functions with C.1 Enter applicable Immediately three CHANNELS inoperable. Conditions and Required Actions QB for the associated DG made inoperable Required Action and by DG - LOVS associated Completion Time instrumentation. not met. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform CHANNEL CHECK. 12 hours SR 3.3.7.2 Perform CHANNEL FUNCTIONAL TEST. 92 days (continued) i 1 l l w.-o e . . su rn , s.,.,,

System 80+ Design ControlDocument RCS Pressure, Temperature, and Flow limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS)

.4.1    RCS Pressure, Temperature, and Flow limits                                                                       . _

_C0 3.4.1 RCS DNB parameters for pressurizer pressure, cold leg temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure 2 [2175 psia) and s [2325 psia),

b. RCS cold leg temperature (T,): _ __

2 [543*F] and 5 [561*F] for < [90%) of RTP, or 2 [550*F] and s [561*F] for 2 [90%) of RTP, and

c. RCS total flow rate 2 [95%) and s [116%) of [445,600]

gpm. (PPLICABILITY: MODES I and 2.

                        --_-------==           . - -
                                                             =-- -NOTE-      --       - - -
                                                                                            ---   = - - - - =

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp in excess of 5% RTP per minute; or

b. THERMAL POWER step in excess of 10% RTP.

                        ------------- =- _ = - - - -           =--------------------- ===- _-------

CTIONS CONDITION REQUIRED ACTION COMPLETION TIME

4. Pressurizer pressure A.1 Restore parameter to 2 hours or RCS total flow rate within limits.

not within limits.

3. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time of Condition A not met.

( ., (continued) Mi J .S ...

                                                                                                                , . , d 
                       , , .       ff eroved Design Meterial- 16.3 Tech Spec                \

page 3.4 1 g- e, a

Svstem 80+ Deslan ControlDocument RCS P/T Limits 3.4.3 i i I i i 1 I I 2400 : I M*

c* :

Hy@vstabc .

($WF,2160psig!

Test : %

                                                                                  !        Q W Ff 2000   -

92PFf Mst seme.

l l2WFf Temperature

l f 1000 -

                                                                            $                                    =

o. :

t' ! (4WF.1290pokll 8 i N 5

             ,,.0   -

i j%y

Re9mtsof ,

         @                                                          :              cm               ' " lr                                 ~

f ,0peradwt /

                                                                  ,I w
                                                   =

S 800 Inchermat' = 5 WFN erFw 5 l ,

fWFN w > no N

Am*mnn j 7 a

                     '"""'*""                                                                                                           I
                                                           . . . . .!                     N t+                                 j .!    w-
                                               =

wFe 4 Wru LsWFx k 4 wa h 0

                                                                                                                                  ?
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v l 0 0 i 50 i i i , , i // s // '/ 100 150 200 250 300 350 400 450 W 550 000 _ IndicatedFluid Temperature, *F - Minimum LTOP disable temperature as a function of heatup rate. Figure 3.4.3-1A RCS Pressure and Temperature Limits (Heatup)

 %nprovedDesign Meteriel 16.3 Tech Spec Page 3.4-8

Ivstem 80+ Desinn Control Document RCS Loops - MODE 3 3.4.5 CTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME , ,

                                                                                                              'ha
                                                                                                      \

47 rd e C. Two required RCS loops C.1 Suspend all. Immediately - ' inoperable. QB operations'. involving a reduction of RCS boron concentration. [4 O3c e, i h, 4 No RCS Loop in MQ optration. _ __ C.2 Initiate action to Immediately restore one RCS loop to OPERABLE status and operation. JRVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l IR 3.4.5.1 Verify required RCS loop is in operation. 12 hours l I iR 3.4.5.2 Verify secondary-side water level in each 12 hours steam generator 2 [25]% wide range indications. iR 3.4.5.3 Verify correct breaker alignment and 7 days indicated power availaole to the required i pump that is not in operation. l l I I aroved Dee> nietwiel.16.3 Tech Spec pp 3.s yg

System 80+ Deslan ControlDocument RCS Loops - MODE 5 (Loops Filled) 3.4.7 i3.4 REACTOR COOLANT SYSTEM (RCS) i

3.4.7 RCS Loops - MODE 5 (Loops Filled) , ,

lLCO 3.4.7 One Shutdown Cooling System (SCS) division shall be OPERABLE I and in operation, and either: l a. One additional SCS division shall be OPERABLE; or

b. The secondary side water level of each Steam Generator l (SG) shall be 2 [25%] wide range indication. __

                 -------_------------ =-
                                                       -NOTES-----------------=    ------

l 1. The SCS pump of the division in operation may be de-l energized for s I hour per 8-hour period provided:

a. No operations are permitted that would cause reduction of the RCS boron concentration; and 1 l b. Core outlet temperature is maintained at least 10*F l below saturation temperature. '

2. One required SCS division may be inoperable for up to 2 hours for surveillance testing provided that the other SCS division is OPERABLE and in operation.

3. No RCP shall be started with one or were of the RCS cold l leg temperatures s [259'F] during cooldown or s the LTOP l disable temperature during heatup (the heatup rate is limited as shown in Figure 3.4.3-1A) unless:

a. Pressurizer water level is < [60]%; or

b. Secondary water temperature of each SG is < [100*F]

above each of the RCS cold leg temperatures. i 4. All SCS divisions may be removed from operation during planned heatup to MODE 4 when at least one RCS loop is in operation.

5. A Containment Spray pump can be manually realigned to meet the requirement of an SCS pump.

i l A- q7 N hPLICABILITY: MODE 5 with RCS loops filled. N dM t - M 4 2/a [ w .,~ ,e., r.~ dA. 1.3-

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I Svst?m 80+ \ p.e i4 k Desi!Ln ControlDocument

g - RCS PIV Leakage i

                                                             ,                 ,.                                                  3.4.13
                                                          )         o r ACTIONS ,'ccr.t u nd)                      \

CONDITION ..., em uoN COMPLETION TIME A. (continued) A.2.1 Isolate the high 72 hours pressure portion of ' the affected system l

                                                                     . ,, from.the low. pressure portion by use of a second closed manual, deactivated automatic or check valve.

9.8 A.2.2 Restore RCS PIV 72 hours leakage to within 1 limit. a

                                                                                               ..                                                    +

Be in N0DE 3.- 6 hours. B. Required. Actions and B.1 ! associated Completion

l Time for. Condition A Ngl

                                                                                           ,s 4.                                                '

not met. - - - B.2 Be in N0DE 5. 16 hours . i i

           'I                         . -                                                                                 -

L 1 I Approved Design Materle!.16.3 Tech Spec page 3.4 31

l Svstem 80+ Deskrn ControlDocument l RCS LEAKAGE Detection Instrumentation j 3.4.14 3.4 REACTOR COOLANT SYSTEM (RCS) , i 3.4.14 RCS LEAKAGE Detection Instrumentation .. .

                                                                                                                                                .l LCO 3.4.14                            The following RCS LEAKAGE detection instrumentation shall be                                                  !

OPERABLE: l

a. Containment sump monitor; and l l

b. Containment atmosphere radioactivity monitor (gaseous or particulate); and __

c. Containment cooler condensate tank monitor.

APPLICABILITY: MODES 1, 2, 3 and 4. l ACTIONS _ i l CONDITION REQUIRED ACTION COMPLETION TIME l i A. Required containment A.1 -------NOTE---------- , sump monitor LCO 3.0.4 is not inoperable. applicable. ; OR I Perform SR 3.4.12.1. Within [4] hours l Required containment ! cooler condensate tank alg} l monitor inoperable. ! l 8151 Once per 24 .: J _ hours -

                                                                                                                   .r:.g q A.2    Restore monitor (s) to       7 days      ~

QjY l OPERABLE status. 2e4

                                                                                                                       , *t a1t e    s (continued)              j.
                                                                                                                             .[m    ,

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l System 80+ oestan c ntrolDocument SIS - Operating 3.5.2 , 1 SURVEILLANCE REQUIREMENTS (continued) l SURVEILLANCE FREQUENCY

                                                                                                                    - .l j

SR 1.5.2.7 Verify each Safety Injection pump starts [18] months automatically on an actual or simulated i actuation signal. w R Verify, by visual inspection, that the [18] months . U).5.2.8 IRWST Holdup Volume Tank is not restricted by debris and trash racks and screens show no evidence of structural distress or abnormal corrosion. y(d-N' l 4 proved Design Meterial- 16.3 Tech Spec page 3.5.g

Syntem 80+ Deslan Control Document ) SIS - Shutdown 4 3.5.3 l 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS) 3.5.3 Safety Injection System (SIS) - Shutdown 4 j . l LCO 3.5.3 Two trains of SIS, one in each division, shall be OPERABLE. APPLICABILITY: MODE 4, 5, and ! 3 MODE 6 with RCS level < [120'-0"] i

4 ACTIONS j CONDITION REQUIRED ACTION COMPLETION TIME l i l i

 . A. Required SIS train                 A.1    Restore required SIS     1 hour

! inoperable. train to OPERABLE

status. ~ l 4 l B. Required Action and B.1.1 Verify RCS level 2 Immediately j associated Completion (120' - 0"]. < Time not met. I j E 5 B.I.2 Initiate actions to Immediately restore RCS level to 2 [120' - 0"]. o E N i

f ----- 24 hours %

                                                                                                       ',        N 9

4' B.2 Reduce RCS temperature to j g N i d < 135* F. (M vg - 3% N } Yt .D' i b i e i < Anoroved Deelen Meteriel 16.3 Tech Spec pay,3,s.7

I System 80+ N&n ControlD*cuma j TRWS S 1 i 3.5. t ACTIONS-(:=....___, e j,r y3 j CONDITION REQUIRED ACTION COMPLETION TIME o$ ! i 4* S *! C. (continued) C.3 Reduce RCS temperature to 24 hours ND

1

A 1

< 135' F.

l AND

C.4 Be in MODE 5. 36 hours .__

c i ,1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify IRWST borated water temperature is 24 hours : within range specified in Figure 3.5.4-1. SR 3.5.4.2 Verify IRWST borated water volume is 7 days 2 [555,800] gallons and s [575,000] gallons. SR 3.5.4.3 Verify that the IRWST Holdup Volume Tank is 7 days [ Empty]. SR 3.5.4.4 Verify IRWST boron concentration is 7 days 2 [4000] ppm and s (4400] ppm. 1 N D I tnY ws 7 sb (g. 9gm_ M p kgertved Denkre Ateneriet 16.3 Tech Spec page 2.5 10

System 80+ Design ControlDocument Containment Air Locks 3.6.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) Verify the OPERABLE 1 hour - door is closed in the . - affected air lock. M A.2 Lock the OPERABLE 24 hours door closed in the affected air lock. .- E A.3 --------NOTE---------- Air lock doors in high radiation areas may be verified locked closed by administrative means. Verify the OPERABLE Once per 31 days door is locked closed in the affected air lock. i B. One or more B.1 --------NOTES--------- containment air locks 1. Required Actions with containment air B.1, B.2, and B.3 lock interlock are not applicable mechanism inoperable. if both doors in the same air lock are inoperable and Condition C is entered.

2. Entry and exit of l containment is permissible under the control of a dedicated

                                          .s               individual.
                                            -j-,j.-----------------------

(continued) l Approved Design Material- 16.3 Tech Spec Page 3.6-3

                                                                                                           .            t System 80+                                                           Design ControlDocumen*

Containment Air Lock 'q ' 3.6.1 -

                                                                                                   ~4     e ACTIONS      m.....    .___,
                                                                                               +-T        j CONDITION                     REQUIRED ACTION              COMPLETION TIME                j di s    r          '

B. (continued) Verify an OPERABLE 1 hour * { door is closed in the affected air lock. Ad : sa 8HD B.2 Lock an OPERABLE door 24 hours closed in the affected air lock. _ AND B.3 --------NOTE---------- Air lock doors in high radiation areas may be verified locked closed by administrative means. l l Verify an OPERABLE Once per 31 days door is locked closed in the affected air lock. C. One or more C.1 Initiate action to Immediately containment air locks evaluate overall inoperaable for containment leakage ! reasons other than rate per LC0 3.6.1. Condition A or B. : 8HD C.2 Verify a door is I hour closed in the affected air lock. AHQ C.3 Restore air lock to 24 hours ! OPERABLE status. l (continued) l l Approved Design Materiel 16.3 Tech Spec Page 3.64 l

Syotem 80+ Design ControlDocument Containment Air Locks 3.6.2 ! C SURVEILLANCE REQUIREMENTS (continued) . iIat t SURVEILLANCE FREQUENCY

                                                                                                          \i.'j SR   3.6.2.2         ---------------------NOTE------------------

Only required to be performed upon entry into cor.tainment. Verify only one door in the air lock can be 184 days opened at a time. l l l i l l l l l l Approved bes> Material.16.3 Toch Spec Page 3.6-6

Sy: tem 80+ Design ControlDocument l l Containment Isolation Valves ' 3.6.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ,(( [,'d$fq ~ A. (continued) A.2 --------NOTE--------- -4nce per 31 :iays l' ' ' for isolation - r s device outside r, Valves and blind . flanges in high containment o , 3 i radiation areas may

i r be verified by use of MQ i g c.,h '

administrative means. i 2 s .'O

                                                                                                                             .' d y
                                              ---------------------               Prior to                            3 entering MODE 4                  -.S      '     o:

Verify the affected U from MODE 5 if '

f penetration flow path not performed is isolated. within the previous 92 days !

for isolation device inside containment B. ---------NOTE--------- B.1 Isolate the affected I hour Only applicable to penetration flow path ' those penetration flow by use of at least paths with two one closed and containment isolation deactivated automatic valves. valve, closed manual valve, or blind One or more penetration flow paths with two containment isolation valves inoperable (except for purge valve leakage and shield building leakage not within limit]. l I (continued) yu

                                                                                     /      't(V ./~'
                                                                                                 ,y        ,1
                                                                                       \     , p  \    -+

l ' (/ ' 0\ ,/ i ApprovedDesign Meterial 16.3 Tech Spor V L g

                                                                                           \      Pope 3.6-8 n ..       e l
                                                                             \      /O

i System 80+ Design controlDocument i Shield Building 3.6.8 SURVEILLANCE REQUIREMENTS (continued) , SURVEILLANCE FREQUENCY l SR 3.6.8.2 Verify shield building structural integrity During shutdown by performing a visual inspection of the for SR 3.6.1.1 exposed interior and exterior surfaces of Type A tests the shield building. SR 3.6.8.3 Verify each Annulus Ventilation System [18] months division with a final flow rate of s ' [18,000) cfm produces a pressure equal to W or more negative than [-0.25] inch water -

                                                                                                                           "    43 gauge / in the annulus within [110 seconds]                                                            W after a start signal.                                                 4                          d h

D N P Z O l l l l l I l l l l Anwoved Deskre Aceterin! 16.3 Tech Spec pp 3 g.gg

System 80+ oestan control Document EFW 3.7.4 ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME E. Required EFW division E.I ---------NOTE--------- inoperable in MODE 4. LC0 3.0.3 and all other LCO Required Actions requiring MODE changes are suspended until one EFW division is restored to OPERABLE -- status. Initiate action to Immediately restore one EFW division to OPERABLE status. SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify each EFW manual, power operated, and 31 days automatic valve in the flow path and in both steam supply flow paths to the steam driven pump, that is not locked, sealed, or otherwise secured in position, is in its correct pasition. 7SR 3.7.4.2 ---------------------NOTE---------------- f-Not required to be performed for the N % 4 . ,-s turbine driven EFW pump until [24] hours after reaching [800] psig in the steam

                                                                                                     'pW   n   r- C generators.

46

                                                                                                         /;
                                                                                                                  '?
                                                                                                                  ..e Verify the developed head of each EFW pump             [31] days on a at the flow test point is 2 the required               STAGGERED TEST developed head.                                        BASIS (continued)

Approved Desigrs Material.16.3 Tech Spec page 3.7 10

Svstem 80+ Deslan ConeelDocument i EFWST 3.7.5 3.7 PLANT SYSTEMS l 3.7.5 Emergency Feedwater Storage Tank-(EFWST) , ,

                                                                                                                           . i<

LCO 3.7.5 Two Emergency Feedwater Storage Tank (EFWST) levels shali be - 2 [350,000] gallons. i APPLICABILITY: MODES 1, 2, and 3. MODE 4 [when steam generator is relied upon for heat removal). ,, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME : A. One EFWST level not A.1 Verify OPERABILITY of 4 hours ! within limit. other EFWST. m : M r Once per 12 .. hours thereafter. A.2 Restore EFWST level 7 days ~

R v 3

to within limit. '

L I .D v<V N s O .

l B. Required Action and B.1 Be in MODE 3. 6 hours ,

associated Completion ; Time not met. M B.2 Be in MODE 4 without [18] hours reliance on steam generator for heat removal. I Anprenaf Denkn nieterW.16.3 Tech Spec page 3.712

System 80+ Deslan ControlDocument UHS 3.7.9 l ACTIONS (continued) CONDITION REQUIRED ACTION COMPLETION TIME C. UHS inoperable in C.1 --------NOTE---- ----- MODES 5 or 6. Enter applicable l Conditions and : Required Actions of LCO 3.7.8 " Station , Service Water System (SSWS)", for SSWS division (s) made - inoperable by the ~sb V" UHS. g Ipitiate Actions to NImmediately ,, gd - OPE E status. go,

, ,' 9

                                                                                                                                       .v
                                                                                                                     ' /)\ a. v.
                                                                                                                               'f D .3
                                                                                                                           \'&\%

4 SURVEILLANCE REQUIREMENTS ! SURVEILLANCE FREQUENCY

 -                                                                                                                                               l l

SR 3.7.9.1 Verify water level of the Ultimate Heat 24 hours ! Sink is 2 [*] [mean sea level). - l SR 3.7.9.2 Verify average water temperature of the 24 hour 7 _ Ultimate Heat Sink is s [95'F]. _ SR 3.7.9.3 Operate each cooling tower fan for 2 [15] 31dayI minutes.

=mw neww 1s.2 recs spec rey,2.7 2s

System 80+ Deslan controlDocument 1

l ADVs 3.7.11 3.7 PLANT SYSTEMS 3.7.11 Atmospheric Dump Valves (ADVs) , , LCO 3.7.11 [Two] ADV lines shall be OPERABLE. APPLICABILITY: MODES 1, 2, and 3. ! [ MODE 4 when steam generator is being relied upon for heat - _ _._. . removal.] .

                                                                                  . _ _ _    ,                         ggAd h c '

ACTION [ 4.g\L CONDITION REQUIRED ACTION COMPLETION TIME i gqd' ' A. One required ADV line A.1 ---------NOTE--------- i inoperable. LCO 3.0.4 is not applicable. Restore ADV line to 7 days OPERABLE status. B. [Two or more required] B.1 Restore [one] ADV 24 hours ADV lines inoperable. line to OPERABLE status, l l C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion - Time not met. -"- C.2 Be in MODE 4, without [12] hours reliance upon steam generator for heat i removal. AND C.3 Be in MODE 5. [24] hours l 4nweved Doets AteserW 16.3 Tech Spec hye 2.7 23 1

f { Svstem 80+ Mn controlDocument l 2 Spent Fuel Assembly Storage , ! 3.7.20 l ! l i3.7 PLANT SYSTEMS n a n o . 3.7.20 Spent Fuel Assembly Storage i { , , f s

                                                                                                                               ,e         r j

LCO 3.7.20 The combination of initial enrichment and burnup of each spent fuel assembly stored in [ Region 2] shall be within tt4 vi\

                                                                                                                    - 'E .3 311 p4
                                                                                                                                              )'1-i                                  acceptable [burnup domain) of Figure 3.7.20-1 [or in                              b ,p                 (V ,

accordance with Specification 4.3.1.1]. e1m-y '+ k ,' f a Whenever any fuel assembly is stored in [ Region 2] of the jAPPLICABILITY: ! fuel storage pool. 3 4 4 C 47 4 (L - a IACTIONS CONDITION REQUIRED ACTION COMPLETION TIME db*j

                                                                                                                      $dg A.       R2quirements of the                        A.1       --------NOTE----------

j LCO not met. LCO 3.0.3 is not _x

EV V >

j applicable. t) ______________________ {' { , , i Initiate action to Immediately I move the noncomplying fuel from [ Region 2]. }s l

                                                                                                                                    \         i IDi     I s

Shf SURVEILLANCE REQUIREMENTS b SURVEILLANCE FREQUENCY 1 SR 3.7.20.1 Verify by administrative means the initial Prior to ! enrichment and burnup of the fuel assembly storing the is in accordance with Figure 3.7.20-1 or fuel assembly Specification 4.3.1.1. in [ Region 2] nap ==r o > mesww. rs.s r.en s,.c p. ,,3. m

Spent Fuel Assembly Storage 3.7.20 40 I I I I I I

35 - - l l 30 - - i-o

y ACCEPTABLE a '

25 - - h_ n. o z ! cc o . [ 20 - - o e I O m 5 15 - -

                                          >                                               UNACCEPTABLE 2$

2 m

                                          <    10   -                                                                             -

i l l t -- . 5 - w . i I I I I I i 0 8 1.5 2.0 2.5 3.0 3.5 4.0 4.5 5.0 5.5 U-235 ENRICHMENT Figure 3.7.20-1 Discharge Burnup vs. Initial Enrichment for Region II Racks i 3YSTEM 80+ 3.7-41 06/17/94-Amendment W j 16.3 Tech Spec r

System 80+ Denkn ControlDocument l AC Sources - Operating 3.8.1 l l 3.8 ELECTRICAL' POWER SYSTEMS 3.8.1 AC Sources - Operating _ ,l

                                                                                                                                       .             .i

, LCO 3.8.1 The following AC Electrical Power Sources shall be OPERABLE. t

a. Two qualified circuits between the offsite transmission network and the onsite Class IE AC Distribution System; and !

b. Two diesel generators (DGs), each capable of supplying one ,

division of the onsite Class IE AC Distribution System. _ ;

c. Automatic load sequencers for Division 1 and Division 2.

t l APPLICABILITY: MODES 1, 2, 3, and 4. j l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required offsite A.1 Perform SR 3.8.1.1 1 hour circuit inoperable. for the required g OPERABLE offsite E o . circuit. E L Once per 8 hours thereafter -

                                                                                                                        '   r * (' Eu I         6 A.2               Declare required
                                                                                                                ~

Dk db 24 hours from discovery of no A Tt u a M feature (s) with no s offsite power offsite power to

4
4 available inoperable when its redundant one train concurrent with d .N7 \%

                                                                                                                       ~ g f %-

l required feature (s) inoperability of - is inoperable. redundant required E { feature (s) (continued) , wear w aneww- se.2 recs sm rose 2.s-1

System 80+ Deslan ControlDocument AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

                                                                                                             ,l
                                                                                                           . 4 ;

A. (continued) A.3 Restore required 72 hours ; offsite circuit to 1 OPERABLE status. MQ l 6 days from discovery of , failure to meet - LC0 l B.1 --------NOTE --- B.1 Perform SR 3.8.1.1 1 hour u Required Action for the OPERABLE - B.3.1 or B.3.2 g required offsite MQ , shall be completed circuit (s). ~ if this Condition Once per 8 hours QI n g is entered. MQ -t thereafter f-e j B.2 Declare required 4 hours from I b One required DG feature (s) supported discovery of d v inoperable, by the inoperable DG Condition B 0 inoperable when its redundant required concurrent with inoperability of v) d feature (s) is redundant inoperable. required feature (s) MD B.3.1 Determine OPERABLE DG [24] hours is not inoperable due to common cause failure. 9.B (continued) l Approved Deelen Metenini- 16.3 Tech Spec Page 3.8-2

System 80+ oesian ControlDocument AC Sources - Operating 3.8.1 ACTIONS CONDITION REQUIREU ACTION COMPLETION TIME i - .i B. (continued) B.3.2 Perform SR 3.8.I.2 24 hours . A for OPERABLE DG. B.4 Verify the combustion 72 hours I turbine generator (CTG) is functional by verifying the CTG -- starts and achieves steady state voltage and frequency within [2] minutes. M B.5 Verify the CTG is 72 hours capable of being - aligned to the ESF M buses associated with the inoperable DG. Once per 8 hours l thereafter M B.6 Restore required DG 14 days to OPERABLE status. M l 15 days from ' discovery of failure to meet LCO l C. Two required offsite C.1 Declare required 12 hours from l circuits inoperable. feature (s) inoperable discovery of y when its redundant 6 Condition C

required feature (s) concurrent with
I is inoperable. inoperability of gO

                                       ~                                   redundant              42 M              l                   required (D

features d' L3 4

+-

l (continued)

                                                                                                   %       3       1 1

Ameved Desfon Meterial.16.3 rech Spec p ,y 3_g.3 l

System 80+ Deslan ControlDocument AC Sources - Shutdown 3.8.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

                                                                                                               .l B.   (continued)                     B.2.2   Suspend movement of        Immediately                          l irradiated fuel                                                 l assemblies.

E B.2.3 Initiate action to Immediately suspend operations with a potential for drM ning the reactor vessel. E B.2.4 Initiate action to Immediately suspend operations involving positive reactivity additions. E B.2.5 Initiate action to Immediately restore required DG(s) to OPERABLE status. C. ----------NOTE--------- C.1 Perform SR 3.8.1.1 for I hour Required Action C.3.1 the OPERABLE required or C.3.2 shall be offsite circuit (s). E completed if this y condition is entered. E g. Once per 8 = hours 9-s One of the two thereafter u required on-site ( gf j sources inoperable.

2 (

(continued) 3 % e O l V % D Approved Design Materies- 16.3 Tech Spec page 3.g.gg

Syst m 80+ Deslan ControlDocument Battery Cell Parameters 3.8.6 Table 3.8.6-1 (Page 1 of 1) ' Battery Electrolyte Requirements , CATEGORY A: CATEGORY B: CATEGORY C: PARAMETER Limits for each Limits for each Allowable value designated pilot connected cell for each cell connected cell

                                                                                                                                               #      l Electrolyte             > Minimum level                      > Minimum level                            Above top of Level                   indication mark, and                 indication mark, and                       plates, and not s 1/4" above maximum                 s 1/4" above maximum                       overflowing level indication                     level indication                                                                      r mark (*)                             mark (*)

Float 2 [2.13] volts 2 [2.13] volts > [2.07] volts Voltage t

                                                                                                                                            > m Specific                2 [1.200]                           2 [1.195]                                   Not more than                      O gravity" 3                                                                                                    [0.020] below the s

4Uk*. n average connected cells Q;O s m -t i Average of all Nd M connected cells 2 [1.195]

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

a. It is acceptable for the electrolyte level to temporarily increase above the specified maximum during equaling charges, provided it is not overflowing.

b. Corrected for electrolyte temperature and level. Level correction is not required, however, when battery charging is < [2] amps when on a float charge.

1

c. Battery charging current is < [2] amperes when on float charge. This is acceptable only during a maximum of [7] days following a battery recharge.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = - - - - _ _ __ _ _ - _ - __ _ _-__-_____________________ I .^ ; .; W Ataseniel.16.3 Tech Spec Page 3.3 36 4

System 80+ Deslan Control Document Inverters - Operating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Inverters - Operating . . LCO 3.8.7 The required Division 1 and Division 2 inverters shall be OPERABLE.

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

One inverter per division may be disconnected from its associated DC bus for 5 24 hours to perform an equalizing charge on its associated battery provided:

a. The associated AC vital bus (es) is energized from its Class IE constant voltage source transformer (s); and

b. All other AC vital buses for both divisions are energized from their associated OPERABLE inverters.

APPLICABILITY: MODES 1, 2, 3, and 4. ACTIONS - CONDITION REQUIRED ACTION COMPLETION TIME A. One required inverter A.1 Power AC vital bus 2 hours inoperable. from its (Class IE] constant voltage source transformer. AND -

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

Enter applicable Conditions and i- -24 hours e. Required Actions of 0- 3'4 1 LCO 3.8.9, l 3 '

                                                           " Distribution                                                                               I

Systems - Operating" V k with any vital bus ,[ .'ej th de-energized. s sj C

                                                                                                                                                    ,zzg , h A.2      Restore inverter to -                                    '

OPERABLE status. B. Required Action and B.1 Be in MODE 3. 6 hours associated Completion Time not met. ANQ B.2 Be in MODE 5. 36 hours i _^ _..= Deafen neeww- 16.2 Tech spec enge 2.s-27

__ _. . .. _ _ _ _ _ . _ _ _ _ _ . . . . . _ _ . ~ .._ __ _ __ . System 80+ Desinn ControlDocument SCS and Coolant Circulation - High Water Level 3.9.4 , 3.9 REFUELING OPERATIONS 3.9.4 Shutdown Cooling System (SCS) and Coolant Circulation - High Water - ' Level . LCO 3.9.4 One SCS division shall be OPERABLE and in operation.

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

The required SCS division may be removed from operation for 5 one hour per (8] hour period, provided no operations are permitted that would cause dilution of the Reactor Coolant _ System boron concentration. APPLICABILITY: MODE 6 with the water level ;t 23 feet above the top of the reactor vessel flange. s ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. SCS division A.1 Suspend operations Immediately requirements not met. involving a reduction ; in reactor coolant l boron concentration. M A.2 Suspend loading Immediately irradiated fuel assemblies in the Core. M Rdd A.3 Initiate action to h -e l. ed'I J - satisfy SCS division .- requirements. L,.J'aM*7 M (continued) ' Anproved Deefpn Nesenini- 16.3 Tech Spee peye 2.g.g

i Syntem 80+ Deslan ControlDocument Design Features 4.0 4.0 Design Features (continued) l 4.3 Fuel Storage l 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with: 4 i

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

[5.0] weight percent; ,

b. K ' s 0.95 if fully flooded with unborated water, O~

wNich includes an allowance f r uncertainties as C/ described in [ Chapter 9 N]; [c. A nominal [9.78] inch center to center distance between fuel assemblies placed in [the high density fuel storage racks];) [d. A nominal [9.78] inch center to center distance between fuel assemblies placed in [the low density fuel storage racks];] [e. New or partially spent fuel assemblies with a discharge burnup in the " acceptable range" of Figure [3.7.20-1] may be allowed unrestricted storage in [either) fuel storage rack (s); and] [f. New or partially spent fuel asse#blies with a discharge burnup in the " unacceptable range" of Figure [3.7.20-1] will be stored in compliance with the [NRC approved procedure, specific approved document, configuration, figure, etc.).] 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 *

[5.0] weight percent; /\ g

b. K s 0.95 if fully flooded with unborated water, h

V wNi'ch includes an allowance for uncertainties as A

                                                                                                  %          {\.

described in (Chapter 9 of CESSAR-DC]; %

c. s 0.98 if moderated by aqueous foam, which K,,,ludes an allowance for uncertainties as described N k',

inc %M'N in [ Chapter 9 ef-6999*R*r]; and A 4D. (continued) wnd outon noenweer.1s.4 Tech spec p ,,,s o.2

  - . .       . . ~ . .            ..-.                            _                 ..     ,

Smm 80 + Deslan controlDocument Design Features 4.0 1 4.0 Design Features (continued) d f\. ' This figure shall consist of [a map of] the site area and provide, as a minimum, the information described in Chapter 2:sMmadielk. % relating to [the map]. g% 9 g*Q l Figure 4.1-1 Site and Exclusion Area Boundaries . Novove<1oenter unter, int.1s.4 Tech spec ,,

System 80+ oeslan controlDocumcnt Reviews and Audits 5.5 4 5.5 Reviews and Audits D 4

q l 5.5.1.1 Functions ( Cow b ve d '3 $: c. Determine whether each item considered under Specifications 1 R l 5.5.1.2.'a through 5.5.1.2.d constitutes an unreviewed safety D question as defined in 10 CFR 50.59; and V (y l; d. Notify the [Vice President - Nuclear Operations] of any '

l safety significant disagreement between the [ review _ organization or individual specified in Specification 5.5.1] ' and the [ Plant Superintendent) within 24 hours. However, the [ Plant Superintendent] shall have responsibility for resolution of such disagreements pursuant to Specification ' 5.1.1. 5.5.1.2 Responsibilities The [ plant review method specified in Specification 5.5.1) shall be used to conduct, as a minimum, reviews of the following:

a. All proposed procedures required by Specification 5.7.1.1 and changes thereto;; b. All proposed programs required by Specification 5.7.2 and
changes thereto; l c. All proposed changes and modifications to unit systems or equipment that affect nuclear safety;: d. All proposed tests and experiments that affect nuclear safety;; e. Review and documentation of judgment concerning prolonged operation with protection channels placed in bypass since the last [ plant review meeting) and the repair of these channels; and; f. All proposed changes to these Technical Specifications (TS),

their Bases, and the Operating License. i5.5.2 I0ffsitel Review and Audit l [The licensee shall describe the provisions for reviews and audits independent of the plant's staff (organization, reporting, and records) and the appropriate ANSI /ANS standards for personnel qualifications. These individuals may be located onsite or offsite provided organizational independence from plant staff is (continued) Approved Design Meteniel- 16.5 Tech Spec page 5.0.y l

System 80+ oeskn controlDocument Reviews and Audits I 5.5 5.5 Reviews and Audits 5.5.3 Records (continued) I

c. An assessment of the safety significance of the review or audit findings; l

l

d. Recommended approval or disapproval of items considered
under Specifications 5.5.1.2.a through 5.5.1.2.f; and l: e. Determination whether each item considered under Specifications 5.5.1.2.a through 5.5.1.2.e constitutes an unreviewed safety question as defined in 10 CFR 50.59.

i l l l l I

                                                                                    .% i U t    64 N

D %) 4.

4- 5 ee-> n 4 v' L- _ _ . _ . . . ,. _ , ,, e I

System 80+ De@n CentrolDrcument Procedures, Programs, and Manuals , 1

c. 5.7 q% l

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                                                                                > m,eQ ,% d c, t y ., . e )                       :

5.7 Procedures, Programs, and Manuals I c., x s, o,</\ (.N !

3 h a 'a

                                                    . - 7ar r i e 6 C e.r)                                                        ,

5.7.2.8 Radiological Environmental Monitoring Program f *l This program is for monitoring the radiation and radionuclides in ~f the environs of the plant. The program shall provide . representative measurements of radioactivity in the highest potential exposure pathways and verification of the accuracy of ; the effluent monitoring program and modeling of environmental l exposure pathways. The program shall be contained in the ODCM, - ] ~ shall conform to the guidance of 10 CFR 50, Appendix I, and shall , include the following: l

a. Monitoring, sampling, analysis, and reporting of radiation l and radionuclides in the environment in accordance with the l methodology and parameters in the ODCM;

b. A Land Use Census to ensure that changes in the use of areas at and beyond the site boundary are identified and that modifications to the monitoring program are made if required -

by the results of this census; and

c. Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

5.7.2.9 Component Cyclic or Transient Limit This program provides controls to track the CESSAR-DC, Chapter 3 cyclic and transient occurrences to ensure that components are maintained within the design limits. 5.7.2.10 Inservice Inspection Program This program provides controls for inservice inspection of ASME I Code Class 1, 2, and 3 components, including applicable supports. The program shall include the following:

a. Provisions that inservice inspection o# ASME Code Class 1, 2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda, as required by 10 CFR 50.55a;

b. Provisions for safety-related snubbers in accordance with 10 CFR 50.55a. The only snubbers excluded from this requirement are installed on nonsafety related systems and then only if their failure, or failure of the system on (continued)

Apprenef Design Aceteriel 16.5 Tech Spec Pope 5.0 21

System 80+ Design ControlDocument Procedures, Programs, and Manuals l 5.7 ' 5.7 Procedures, Programs, and Manuals 5.7.2.13 Secondary Water Chemistry (continued)

f. A procedu'/e identifying the authority responsible for the f interprr,tation of the data and the sequence and timing of 9 Ju admini.strative events, which is required to initiate cor:ective action.

r N f Q j s

                                                                                             *N            A ..
                                                                                                 . 2      .

5.7.2.14 Ventilation Filter Testing Program (VFTP) 3 r; A program shall be established to implement the following

                                                                                                ]   + @*

(\ required testing of Engineered Safety Feature (ESF) filter 7 2 ventilation systems at the frequencies specified in [ Regulatory Guide 1.52], and in accordance with [ Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified below [i 10%). a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < [0.05]% when tested in accordance with [ Regulatory Guide 1.52, Revision 2, and ASME N510-1989, at the system flowrate specified as follows [i 10%): ESF Ventilation System Flowrate Subsphere Building Fuel Building Exhaust

Control Complex Annulus

b. Demonstrate for the Control Complex Ventilation System that an inplace test of the charcoal absorber shows a penetration and system bypass < [0.5]% when tested in accordance with

[ Regulatory Guide 1.52, Revision 2, and ASME N510-1989] at the system flowrate specified as follows [i 10%): l 1 Flowrate l l Control Complex

1 Values to be determined by system detail design (continued) ppprend Design Material- 16.5 Tech Spec Page 5.0-25

i l Syntem 80+ oesten control Document Procedures, Programs, and Manual

i. 5.

5.7 Procedures, Programs, and Manuals 1 3 T %

7

l 5.7.2.14 Ventilation Filter Testing Program (VFTP) ( e. , d ' W M > % * ! c. Demonstrate for the Control Complex Ventilation System tha 1Y k a laboratory test of a sample of the charcoal absorber, wh M D b-I l obtained as described in [ Regulatory Guide 1.52, Revision j 2], shows the methyl iodide penetration less than the valu.

specified below when tested in accordance with [ ASTM D3803- ,

1989) at a temperature of s [30*C] and greater than or equal to the relative humidity (RH) specified as follows: -- l Penetration RH i Control Complex *

i i d. For each of the ESF systems, demonstrate the pressure drop across the combined HEPA filters, the prefilters, and the .

1 charcoal absorbers is less than the value specified below i when tested in accordance with [ Regulatory Guide 1.52, i Revision 2, and ASME N510-1989] at the system flowrate

specified as follows [i 10%)

4 l ESF Ventilation System Delta P Flowrate j Subsphere Building i Fuel Building Exhaust *

l Control Complex Annulus 1 3

! e. Demonstrate that the heaters for each of the ESF systems i dissipate the following specified value [i 10%] when tested , l in accordance with [ASME N510-1989]: i ESF Ventilation System Wattage ' , Subsphere Building ~ i Fuel Building Exhaust

l Control Complex i Annulus i The provisions of SR 3.0.2 and SR 3.0.3 are applicable to

,                      the VFTP test frequencies.

i* Values to be determined by system detail design. (continued) w one6,, ueenew- ss.s roch spec rey, s.o.x

!                                                                                                                       e

t System 80+ oeskn controlDocument oce es, a.s, and Manuals a s 5.7 k

                                                                                                           /            f
   ,5.7 4;g     M#                                   [

Procedures, Programs, and Manuals f e s ,,J_- g g 'h

    .Mwx-50-s,An=g e m g                                             m%Lw                                    y\,'y
     .7.2.15                Explosive Gas and Storage Tank Radioactivity Monitoring Pr           ram T s progr       provides control for       nti     y explosive gas            %           .

x ures ont ne t g (Was Gas Ho p System], [the quantity ' ,, z \ of r activit containe M gas storage tanks or fed into the %o offgas treatment system, and the quantity of radioactivity i contained in unprotected outdoor liquid storage tanks). The gaseous radioactivity quantities shall be determined following the . methodology in [ Branch Technical Position (BTP) ETSB 11-5,

                            " Postulated Radioactive Release due to Waste Gas System Leak or Failure"]. The11guldradwastequantitiesshallbedetermind.1in accordance with [ Standard Review Plan, Section 15.7.3, " Postulated Radioactive Release due to Tank Failures").

The program shall include:

a. The limits for the concentrations of hydrogen and oxygen in the [ Waste Gas Holdup System] and a surveillance program to -

ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b. A surveillance program to ensure that the quantity of radioactivity contained in [each gas storage tank and fed into the offgas treatment system] is less than the amount that would result in a whole body exposure of 2 0.5 rem to any individual in an unrestricted area, in the event of [an uncontrolled release of the tanks' contents); and

c. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the

[l.iquidRadwasteTreatmentSystem)islessthantheamount that would result in concentrations less than the limits of 10 CFR Part 20, Appendix S, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an I uncontrolled release of the tanks' contents. l l The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. (continued) Appmed Deenpo na nder- 16.6 Teek Seen Pese 8.0 27 s a it:et *661 et tt tot! 3MN WOd3

Syntem 80+ i \ oestan contrat Document Procedures, Programs, and Manuals 5.7 Procedures, Programs,andManuals(@ g yedh 4- f-fai mX f.1, ( -c k C A /d ' 5.7.2.16 Diesel Fuel Oil Testing Program + ' A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.. The /4 flL program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM standards. The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to

storage tanks by determining that the fuel oil has:

1. An API gravity or an absolute specific gravity within limits;

2. A flash point and kinematic viscosity within limits for ASTM 20 fuel oil;

3. A clear and bright appearance with proper color.

b. Other properties for ASTM 2D fuel oil are within limits within 30 days following sampling and addition to storage tanks.

c. Total particulate concentration of the fuel oil is [<10 mg/t] when tested every 31 days in accordance with ASTM D-2276, Method A-2 or A-3.

5.7.2.17 Fire Protection Program This program provides controls to ensure that appropriate fire protection measures are maintained to protect the plant from fire and to ensure the capability to achieve and maintain safe shutdown in the event of a fire is maintained. 5.7.2.18 Common Mode Failure Evaluation Program This program provides controls to ensure that appropriate software and hardware evaluation procedures, to protect the plant from common mode failure, are established to ensure that redundant system capability is not adversely affected. This program shall evaluate the cause of the inoperability, the affected components, and the plans and schedule for completing proposed remedial actions. If a determination is made that a common mode failure exists within independent channels or independent systems credited to provide functions controlled by Technical Specifications, then a Special Report shall be submitted in accordance with Specification [5.9.2.f]. T>, c+ o n c "" P"Y* *"* p y, (_ M m e it? / h c 7* '" "5 h

                                                         ,-f- M f.c % 6 L '

W Du** ^****w- 1s.s rock spec rose s.o.2s

Sv' tem 80+ Deslan ControlDocument Reporting Requirements 5.9 5.0 ADMINISTRATIVE CONTROLS 5.9 Reporting Requirements _. .

                                                                                                  . 4 5.9.1          Routine Reoorts The following reports shall be submitted in accordance with 10 CFR 50.4.

G.9.1.1 Startup Report - l A summary report of plant startup and power escalation testing shall be submitted following:

a. Receipt of an Operating License;

b. Amendment to the license involving a planned increase in power level;

c. Installation of fuel that has a different design or has been manufactured by a different fuel supplier; and

d. Modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit, [

g i The initial Startup Report shall address each of the startup tests 4 identified in 1999WBE Chapter 14, and shall include a 5 description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any

                                                                                            &n b

N corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific Q'RJ details required in license conditions based on other commitments shall be included in this report. Subsequent Startup Reports shall address startup tests that are necessary to demonstrate the ' acceptability of changes and modifications. Startup Reports shall be submitted within 90 days following completion of the Startup Test Program; 90 days following resumption or commencement of commercial power operation; or 9 months following initial criticality, whichever is earliest. If i the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation), supplementary reports shall 1 be submitted at least every 3 months until all three events have been completed. (continued) AmomeDeske meterw 1s.s rech spec reye s.o.2s

System 80+ Deshan ControlDocument Report.ing. Requirements

                                                                                                                ' N'5. 9

[ r,v' N j qI (

5.9 Reporting Requirements

                                                   -   b b f i Y.
                                                                          , r. .n ,_ g/ f.
                                                                                        -          ,     u               h.j,j'               _
                                                                                                                                             ,'            vs Q   E*Q Annual 5.9.1.2 e # Reports   mal 6Qr M .                            I ' I ,s
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                                                                                                                                               .                    i N~ .,_ _                                                                                                                    /, 4

______________________________ NOTE - - - - - - - - - - - - - - - - - - - - - - - - - A single submittal may be made for a multiple unit station. r'3 - f-The submittal should combine sections common to all units at the station. 4 g f g s s y cC *t i Annual Reports covering the activities of the unit as described - ,t below for the previous calendar year shall be submitted by March 31 1 e) e of each year. [The initial report shall be submitted by March 31 of g .h the year following initial criticality.] 'v u n

Reports required on an annual basis include: Q f'

a. Occupational Radiation Exposure Report i The tabulation on an annual basis of the number of station, utility, and other personnel (including contractors) receiving

' exposures > 100 mres/yr and their associated man rem exposure according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance [ describe maintenance], waste processing, and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. l Small exposures totalling < 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of I the total whole body dose received from external sources l should be assigned to specific major work functions; and l [b. Any other unit unique reports required on an annual basis.] 5.9.1.3 Annual Radiological Environmental Operating Report

                    ------------------------------NOTE--------------------------------                                         I A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station.

The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for (continued) 4prownf Deefpn nieseniel 16.5 Tech Spec Pope 5.0-22

System 80+ Design C*ntrol Document Reporting Requirements 5.9 5.9 Reporting Requirements 4 - 5.9.2 Special Reoorts (continued) R,]

c.

When a Special Fe pr;& of $ /\ , LCO 3.3.11, " PostReport is Monitoring Accident required by Condition B,hion Instrumenta (PAMI)," a report shall be submitted within 14 days. The report shall outline the preplanned alternate method of dN

                                                                                                        ' =y <3 Q.

i monitoring, the cause of the inoperability, and the plans 4 A, and schedule for restoring the instrumentation channels of 9 the function to OPERABLE status. S I

d. Following each inservice inspection of steam generator (SG) tubes, in accordance with the SG Tube Surveillance Program, the number of tubes plugged and tubes sleeved in each SG shall be reported to the NRC within 15 days.

The complete results of the SG tube inservice inspection shall be submitted to the NRC within 12 months following the completion of the inspection. The report shall include:

1. Number and extent of tubes inspected,

2. Location and percent of wall-thickness penetration for each indication of an imperfection, and

3. Identification of tubes plugged and tubes sleeved.

Results of SG tube inspections which fall into Category C-3 shall be reported to the NRC prior to resumption of plant operation. This report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

e. When a special report is required by Section 5.7.2.18, i " Common Mode Failure Program," a report shall be submitted I

l to the NRC within 30 days. The report shall include a description of the cause of the failure, the affected components, and plans and schedule for completing proposed remedial activities. I i l ApprovedDeeen Atatedel- 16.5 Tech Spec Pege 5.0 36

! Sveteen 80+ Denhm C=-arolDocument (High Radiation Area) , 5.11 [5.11 Hig gh 1 - \.- 1 4M ftd w cb Y N ~- . 11.2 igh Radiation Areas ith Dose Rates Greater th n 1.0 res/ hour,

but less than 500 r s/ hour:** ( r_ ~4 rei , _ ((

~

                               . Acces        o, and activities in, each such area shall be

i con lled by means of an RWP or equivalent that includes i- cification of radiation dose rates in the immediate work I i rea(s) and other appropriate radiation protection equipment l and measures. ' "
i t

! c Individuals qualified in radiation protection procedures may j be exempted from the requirement for an RWP or equivalent ! while wrforming radiation surveys in such areas provided that t my are following plant radiation protection { procedures for entry to, exit from, and work in such areas. i d. Each individual (whether alone or in a group) entering such

an area shall possess: : -

i 1 (1) An alarming dostmeter with an appropriate alarm ! setpoint, or 4 (2) A radiation monitoring device that continuously transmits dose rates and cumulative dose information to a remote receiver monitored by radiation protection a personnel responsible for controlling personnel . radiation exposure within the area with the means to ' communicate with and control every individual in the area, or (3) A self-reading dosimeter and, 1 (i) Se under the survsillance, as specified in the RWP or equivalent, of an individual qualified in radiation protectiu procedures, equipped with a radiation monitoring and indicating device who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RUP or equivalent, by mean,s of closed circuit television, of personnel qualified in radiation ; protection procedures, responsible for ' controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area. (contir,ued) wneeenannee,w. re.s reek seen n oe s.est

                .a                                      risei            *6si et ti                                                        tosi aus wona

Svetent 80+ _ neeeen ControlDocument (High Radiation A [5.11 High Radiation Area] M wi,ea 7 -

                                                                                                    .i
                                                                                                           /

W*SW !$) f : 5.11.2 igh Radiation Areas witit. f e Rates e ter t 1.0 res/ hour,

q

                                                                                                         ^
                                                                                                       ~

but less than 500 ra /Nour:** (continued) i

e. Entry into su areas shall be made only after dose rates in I the area been determined and entry personnel are knowledge le of them.

9 7# G

f. Sc c

dividual areas that are wtthin a larger area that is rolled as a high radiation area, where no enclosure e ists for surpose of locking and where no enclosure can essonably se constructed around the individual area need not be controlled by a locked door or gate, but shall be barricaded, conspicuously posted as a high radiation area, and marked by a conspicuous flashing li ht activated at the area as a warning device which is clearY.y visible from all

access points to the area.

At 30 centimeters surface penetrated by(12 inches)iation.from the rad the radiation source or from any At 1 meter from the radiation source or from eny surface penetrated by

the radiation. I l 1 ) ) 1 l 1 a L Ann,mm aanww. te.s rees spee t rare s.on , s .a cirei essi et it sati aan n aa

                                     -                                                                                                              l 1

System 80+ Design ControlDocument

                                                                                                          - Reactor Core SLs B 2.1.1                 i t

i BASES i i

       '                                                                                           ^                              ~ ~ - -

I SAFETY LIMIT 2.275 l VIOLATIONS (continued) If SL 2.1.1.1 or SL 2.1.1.2 is violated, a Licensee Event Report shall be prepared and submitted within 30 days to the

   ~ - .  ..-         .. .. .... . . RRC, the senior samgement of the nrclear-plant, .and the _~.                                              - ,

utility Vice President - Nuclear Operations. This I requirement is in accordance with 10 CFR 50.73 (Ref. 4). l 2.2,6 If SL 2.1.1:1 or SL 2.1.1.2 is violated, restart cf the unit shall not commence 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. 10 CFR 50, Aspendix.A,2GDC 10. f.c* - REFERENCES 1. -

                                                                                                          =         -e { k{Ah_',

c. w..,- ., C 1ipters?.A and O - 15.

                                                                                                          -y                         .-.3

3. 10- CFR 50.72. h8- D d

4. 10 CFR 50.73. I

       ^1
                              . -                                                                             :                                r_

Approved Design Material. on . . l 5pec Beses Page B 2.0-5

System 80+ Deslan controlDocument RCS Pressure SL B 2.1.2 BASES (continued) SAFETY LIMIT 2,2,6 *

VIOLATIONS (continued) If the RCS pressure SL is violated, restart of the unit shall not commence 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.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 14, GDC 15, and GDC 28.

2. ASME, Boiler and Pressure Vessel Code, Section III, Article NB-7000.

3. ASME, Boiler and Pressure Vessel code, Section XI, Article IWX-5000. V 4.

Q 10 CFR 100. ' 4 5. 6. CC';';",", OQ Chapter 5. +

[ASME, USAS B31.1, Standard Code for Pressure Piping).

}(

7. 10 CFR 50.72. h0

8. 10 CFR 50.73.

AMwesed Denkm Meteniel- 16A Tech Spec Betas pay, g 2.0.g0

System 80+ ~->-- Control Documerrt i LCO Applicability B 3.0 3.0 LCO APPLICABILITY

LCO 3.0.2 ACTIONS.) The second type of Required Action specifies the * * (continued) remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation. Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications. _ The nature of some Required Actions of some Conditions l necessitates that, once the Condition is entered, the Required Actions must be completed evea though the

associated Conditions no longer exist. The individual LCO's l ACTIONS specify the Required Actions where this is the case.

i An example of this is in LCO 3.8.I, "AC Sources - ' Operating." l The Completion Times of the Required Actions are also - applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance,. corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for-operational convenience. Alternatives that would not result in redundant equipment being inoperable should be used instead. Doing so limits the time both divisions / subsystems / trains of a safety function are inoperable and limits the time other conditions exist which result in LCO 3.0.3 being entered. Individual Specifications may specify a time limit for performing an SR when equipment is rescued from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or - bypassed. When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition (s) are entered. (continued) Anume on+, areeuw- 1sA rech see sua revs 2.k 2 %D# 0

System 80+ oesian control Document SDM B 3.1.1 1 BASES (continued) SURVEILLANCE SR 3.1.1.1 (continued) REQUIREMENTS

e. Xenon concentration;

f. Samarium concentration; and

g. Isothermal temperature coefficient (ITC).

Using the ITC accounts for Doppler reactivity in this calculation because the reactor is suberitical, and the fuel temperature will be changing at the same rate as that of the RCS. The Frequency of 24 hours is based on the generally slow change in required boron concentration, and it also allows sufficient time for the operator to collect the required data, which includes performing a boron concentration - analysis, and complete the calculation.

\]

l REFERENCES 1. 10 ffR 50, Appendix A, GDC 26. /2h

2. -CE55AR-DCg Chapter 15 :d I

3. 10 CFR 100. C 1

 ^;;-.J Design Meteria!.16A Tech Spec Bones                                        pay, y g, p.7 l

System 80+ oesion contrat Document RTCB B 3.1.2 : BASES (continued) ACTIONS Ad If any reactor trip circuit breakers are closed they are to be immediately opened to ensure that there can be no inadvertent CEA withdrawal. SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Verification of the reactor trip circuit breakers being open ensures that the CEAs cannot be inadvertently withdrawn. The Frequency of [12] hours is based on engineering judgment and is considered adequate since normal procedural controls would make the unauthorized or inadvertent closing of reactor trip circuit breakers unlikely. ge; % , A REFERENCES 1. jES';A" DQ Appendix 19.8A, Shutdown Risk Evaluation

                                .. ;;;r. .

71 jp ,4 1 y i s

2. N Chapter 16, Technical Specifications, LCO - y> j 3.1.1.

l i 4 l } 4preved W Meterial- 16A Tech Spec Rosee page g 2.19

A System 80+ Omslan ControlDocument Reactivity Balance

l B 3.1.3 I

l BASES I l ACTIONS JL1 (continued) s Time is reasonable, based on operating experience, for i reaching MODE 3 from full power conditions in an orderly j manner and without challenging plant systems. i ! SURVEILLANCE SR 3.1.3.1 ' ! -REQUIREMENTS . i Core reactivity is verified by periodic comparisons of ! measured and predicted RCS boron concentrations. The comparison is made considering that other core conditions

;                                are fixed or stable including CEA position, moderator

temperature, fuel temperature, fuel depletion, xenon

!                                concentration, and samarium concentration. The Surveillance l                                 is performed prior to entering MODE I as an initial check on                         -

i core conditions and design calculations at BOC. The SR is i ] modified by three Notes. The first Note indicates that the i normalization of predicted core reactivity to the measured i j value must take place within the first 60 effective full l l power days (EFPD) after each fuel loading. This allows l i sufficient time for core conditions to reach steady state, l 1 but prevents operation for a large fraction of the fuel l cycle without establishing a benchmark for the design ! calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering M0DE 1, is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (e.g., QPTR) for prompt indication of an anomaly. A Note, "Only required after 60 EFPD," is added to the Frequency column to allow this. Anotner Note indicates that the performance of SR 3.1.3.1 1 is not required prior to entering MODE 2. This Note is ! required to allow a MODE 2. entry to verify core reactivity ~ because Applicability is for. MODES I and 2. REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29. '

2. CESS M OCu Chapter 15.

                                                                                                           ~
    . .avenf Deelyn Matente!- 16A Tech Spec Basee                                            Page B 3.1 15

j

!     System 80+                                                        oe-tan control Document MTC

, B 3.1.4 4 1 BASES l SURVEILLANCE SR 3.1.4.1 and SR 3.1.4 2 (continued) i REQUIREMENTS SR 3.1.4.2 is modified by a Note that indicates performance

is not required prior to entering MODE 1 or 2. Although

! this Surveillance is applicable in MODES I and 2, the reactor must be critical before the Surveillance can be

completed. Therefore, entry into the applicable MODE prior i

to accomplishing the Surveillance is necessary. l SR 3.1.4.2 is modified by a second Note that indicates, if extrapolated MTC is more negative than the E0C COLR limit, l the Surveillance may be repeated, and that shutdown must 1 occur prior to exceeding the minimum allowable boron j concentration at which MTC is projected to exceed the lower limit. An engineering evaluation is performed if the j extrapolated value of MTC exceeds the Specification 1* .i-REFERENCES 1. 10 CFR 50, Appendix A, GDC 11.

2. 't.65aR-D% Chapter 15. > _ _

i 1 i ) i 1 l AMweved Deelyn Metene!- 16A Tech Spec Bases Page B 3.120

System 80+ Dorian Control Document i CEA Alignment B 3.1.5 BASES

SURVEILLANCE SR 3.1.5 1

, REQUIREMENTS (continued) Verification of full strength CEA drop times determines that the maximum CEA drop time permitted is consistent with the assumed drop time used in the safety analysis (Ref. 3). Measuring drop times prior to reactor criticality, after I reactor vessel head removal, ensures the reactor internals and CEDM will not interfere with CEA motion or drop time, - and that no degradation in these systems has occurred that would adversely affect CEA motion or drop time. Individual CEAs whose drop times are greater than safety analysis assumptions are not OPERABLE. This SR is performed prior to i criticality due to the plant conditions needed to perform the SR and the potential for an unplanned plant transient if the Surveillance were performed with the reactor at power. 1

REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26.

i

,                        2.      10 CFR 50.46.

4 i 3. -CE55AR-egg Chapter 15. _; , 4 .I I l 4 i 4 AnweredDeekn Mateda!.16A Tech Spec Basee rege g 2. g.2y

Svstem 80+ Deslan ControlDocument Shtudown CEA Insertion Limits B 3.1.6 BASES

SURVEILLANCE SR 3.1.6.1 (continued) REQUIREMENTS Since the shutdown CEAs are-positioned manually by the control room operator, verification of shutdown CEA position at a Frequency of 12 hours is adequate to ensure that the shutdown CEAs are within their insertion limits. Also, the Frequency takes into account other information available to < the operator in the control room for the purpose of - monitoring the scatus of the shutdown CEAs. REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26.

2. 10 CFR 50.46.

3. CESSC-Z , Chapter 15.

_q _ AnementDen6n stedet- 16A Teen spec suas pope s 2.120

4 Syst~m 80+ Design ControlDocument

l. legula' ting CEA Insertion Limits

B 3.1.7 s /< 4 N -

5 ! BASES {t$hti'ahedg - j - :- - v BACKGROUND regulat i limits ensure the required SDM is 1 l (continued) maintai i l i.- Operati ..

                                                                         .ect LC0 limits will prevent fuel                    - - - '
                             . . _, . cladding f ailures thatwouio 'ore'attrthe primary fisvion-" -

product barrier and release fission products to the reactor i

coolent in the event of a LOCA, loss of flow, ejected CEA, 4

or ott.9r accident requiring termination by a Reactor j Protection System trip function. ' i l

APPLICABLE The fuel clidding must not sustain damage as a result of 4 SAFETY ANALYSES normal operation (Condition I) and anticipated operational.

i occurrences (Condition II). The acceptance criteria.for the i

~ ~

regulating CEA insertion, part strength CEA insertion, ASI, i - and T, LCOs preclude core poder distributions from. occurring that would riolate.th,e folicwing fuel design criteria:

                                          ~                                                                         '
         ~

a. Durini a larg'e' bee'ak LOCA',bhe peak cladding ~ l ter.t a ature must' not (x:eedia limit of 2200!E,+# !

10 CFt 50~~46 (Ref. f);

                                                               .                       . '.                                              l

b. Durin; a l'ss o of' forced reactor coolant flow accident, i there must be at least a 95%. probability at a 95% - )

confidence level (the 95/95 DNB criterion) that the i hot fuel rod in.the core does not experience a DNB l condition;

c .- During an ejected CEA accident, the fission energy input to the fuel must not exceed 280 cal /gm (Ref.'3);

and ; 1

d. The CEAs must be capable of shutting down the reactor with a minimum required SDM, with the highest worth .

CEA stuck fully withdrawn, GDC 26 (Ref.1). l Regulating CEA position, ASI, and T -are process variables that together characterize and contEol the-three dimensional power distrit ution of the reactor core. Fuel cladding damage does not occur when the core is operated ou u.ide these LCOs during normal operation. . However, f u el cladding damage could result should an accident oc n:r with simultaneous violation of one or more of

                                                                                                        .   (continued) m               ee Approved Desian & *~= ~' *ra ' -h Spec Ha >                                                           Page B 3.139

i j System 80+ D stan controt Document ! Regulating CEA Insertion Limits B 3.1 ~

BASES M e l

APPLICABLE these LCOs. Changes in the power distribution can cause _D48 k N [ i SAFETY ANALYSES increased power peaking and corresponding increased local .' (continued) LHRs. l The SDM requirement is ensured by limiting the regulating and shutdown CEA insertion limits, so that the allowable } inserted worth of the CEAs is such that sufficient

reactivity is available in the CEAs to shut down the reactor --

! to hot zero power with a reactivity margin that assumes the j maximum worth CEA remains fully withdrawn upon trip (Ref. j 3.) 4 j Operation at the insertion or ASI limits may approach the

maximum allowable linear heat generation rate or peaking i present. Operation at the j factor, with insertion the limit allowed may also inT,dicate the maximum ejected CEA j worth could be equal to the limiting value in fuel cycles _

that have sufficiently high ejected CEA worths. The regulating and shutdown CEA insertion limits ensure that

i. safety analyses assumptions for reactivity insertion rate,

SDM, ejected CEA worth, and power distribution peaking j , factors are preserved (Ref. 3).

1 j The regulating CEA insertion limits satisfy Criterion m of

the NRC Policy Statement.

I jlCO The limits on regulating CEA sequence, overlap, and physical : ! insertion, as defined in the COLR, must be maintained i because they serve the function of preserving power distribution, ensuring that the SDM is maintained, ensuring ,

that ejected CEA worth is maintained, and ensuring adequate negative reactivity insertion on trip. The overlap between '

i regulating banks provides more uniform rates of reactivity l insertion and withdrawal, and is imposed to maintain i acceptable power peaking during regulating CEA motion. The j power dependent insertion limit (PDIL) alarm circuit is , j required to be OPERABLE for notification that the CEAs are outside the required insertion limits. When the PDIL alarm circuit is inoperable, the verification of CEA positions is 1 increased to ensure improper CEA alignment is identified ] before unacceptable flux distribution occurs. I (continued) [w.:auw,a**ww saa run seu sua r,,, a a. se i

l System 80+ Desion controlDocump Regulating CEA Insertion Lino 6s j B 3.1.7 1 4 BASES N g-ACTIONS B.1 and B.2

(continued) N If the CEAs are inserted between the long term steady state ,, '

h insertion limits, the transient insertion limits for intervals > 4 hours per 24 hour period, and the short term f steady state insertion limits are exceeded, peaking factors . l can develop that are of immediate concern (Ref. 3). l ~ Additionally, since the CEAs can be in this condition without misalignment, penalty factors are not inserted in the core protection calculators (CPCs) to compensate for the I developing peaking factors. Verifying the short term steady ; state insertion limits are not exceeded ensures that the 1 i peaking factors that do develop are within those allowed for continued operation. Fifteen minutes provides adequate time for the operator to verify if the short term steady state insertion limits are exceeded. - Experience has shown that rapid power increases in areas of the core, in which the flux has been depressed, can result in fuel damage as the LHR in those areas rapidly increases. Restricting the rate of THERMAL POWER increases to s 5% RTP , per hour, following CEA insertion beyond the long term steady state insertion limits, ensures the power transients experienced by the fuel will not result in fuel failure (Ref. 3). L.1 With the regulating CEAs inserted between the long term steady state insertion limit and the transient insertion limit, and with the core approaching the 5 effective full power days (EFPD) per 30 EFPD, or 14 EFPD per 365 EFPD limits, the core approaches the acceptable limits placed on operation with flux patterns outside those assumed in the long term burnup assumptions. In this case, the CEAs must be returned to within the long term steady state insertion limits, or the core must be placed in a condition in which the abnormal fuel burnup cannot continue. A Completion Time of 2 hours is a reasonable time to return the CEAs to within the long term steady state insertion limits. (continued) AppmvedDeelges Atatend 16A Tech Spec Bases page g 3. g.42

System 80+ O~sion control Document Regulating CEA Insertion Limi+e 8 3.1 BASES W d 4 ACTIONS .C.d (continued) The required Completion Time of 2 h:urs from initial discovery of a regulating.CEA group outside the limits unt its restoration to within the long term steady state limit shown on the figures in the COLR, allows sufficient time io, borated water to enter the Reactor Coolant System from the chemical addition and makeup systems, and to cause the - regulating CEAs to withdraw to the acceptable region. It is reasonable to continue operation for 2 hours after it is discovered that the 5 day or 14 day EFPD limit has been exceeded. This C mpletion Time is based on limiting the potential xenon r, 'istribution, the low probability of an accident, and the s6aps required to complete the action. D.1.1. D.I.2. D.2.1. and D.2.2 - t If the regulating CEA insertion limits are not met, then SDM must be verified by performing a reactivity balance calculation, considering the effects in SR 3.1.1.1 Bases. One hour is sufficient time for conducting the calculation l

                   ,                     and commencing boration if the SDM is not within limits.                                           ;

With the Core Operating Limit Supervisory System out of service, operation beyond the short term steady state insertion limits can result in peaking factors that could approach the DNB or local power density trip setpoints. Eliminating this condition within 2 hours limits the magnitude of the peaking factors to acceptable levels i l (Ref. 3). Restoring the CEAs to within the limit or reducing THERMAL POWER to that fraction of RTP that is allowed by CEA group position, using the limits specified in the COLR, ensures acceptable peaking factors are maintained. L.1 With the PDIL circuit inoperable, performing SR 3.1.7.1 within 1 hour and every 4 hours thereafter ensures improper CEA alignments are identified before unacceptable flux distributions occur. (continued) _c__w menau. tea ras seu num rene a a s es

System 80+ D*sion Control Document Regulating CEA Insertion l BASES (/rSttAihed) t 4 . 1 ACTIONS f1 (continued) When a Required Action cannot be completed within the required Completion Time, a controlled shutdown should t commenced. The allowed Completion Time of 6 hours is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems. -- 1 i SURVEILLANCE SR 3.1.7.1 REQUIREMENTS With the PDIL alarm circuit OPERABLE, verification of each l regulating CEA group position every 12 hours is sufficient i to detect CEA positions that may approach the acceptable ' limits, and provide the operator with time to undertake the - Required Action (s) should the sequence or insertion limits be found to be exceeded. The 12 hour Frequency also takes into account the indication provided by the PDIL alarm circuit and other information about CEA group positions available to the operator in the control room. SR 3.1.7.1 is modified by a Note indicating that entry is allowed into MODE 2 without having performed the SR. This is necessary, since the unit must be in the applicable MODES in order to perform Surveillances that demonstrate the LCO limits are met. ' SR 3.1.7.2 ' Verification of the accumulated time of CEA group insertion between the long term steady state insertion limits -and the transient insertion limits ensures the cumulative time limits are not exceeded. The 24 hour Frequency ensures the ; operator identifies a time limit that is being approached before it is reached. I SR 3.1.7.3 Demonstrating the PDIL alarm circuit OPERABLE verifies that i DIL alarm circuit is functional. The 31 day Frequency takes into account other Surveillances being performed at shorter Frequencies that identify improper CEA alignments. (continued) wona> ***ww 1as run spec sua roue n 2.1.a

System 80+ Decian control Document Regulating CEA Insertion Limits B 3.1.7 , 1 BASES (continued) ! REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26. 1

2. 10 CFR 50.46. l

3. E55an- A Chapter 15 l A -

N l l I l l

           .                                                                                                         I l

i l l l

Weed Dennon Atatental- 16A Tech Spec Bases pay, g 2, g.45

I 1 System 80+ Deslan Control Document Part Strength CEA Insertion Limits B 3.1.8 BASES (continued) t REFERENCES 1. 10 CFR 50, Appendix A, GDC 10 and GDC 26. l

2. 10 CFR 50.46.

3. -CE55AR-DQ Chapter 15. 7, ,

l 1

i l 'l I I

l i l l l I i l I i 1 t Anwoved Design Metend 16A Tech Spec Bases y,y, g y g.go

l l System 80+ - Destan contrar Document STE - SDM B 3.1.9 , , BASES - l ACTIONS Ad (continued). l must be accomplished by increasing the RCS boron concentration. The required Completion Time of 15 minutes - for initiating boration allows.the operator sufficient time to align the valves and start the boric acid pumps and is consistent with the Completion Time of LCO 3.1.1. SURVEILLANCE SR 3.1.9.1 REQUIREMENTS Verification of the ~ usition of each partially or fully withdrawn full strength or part strength CEA is necessary to ensure that the minimum negative reactivity requirements for insertion on.a trip are preserved. A 2 hour Frequency is sufficient for the operator to verify that each CEA position - l 1s within the acceptance criteria. ' SR 3.1.9.2 i Prior demonstration that each CEA to be withdrawn from the i

                 ,               core during PHYSICS TESTS is capable of full insertion, when                            !

tripped from at least a 50% withdrawn position, ensures that j the CEA will insert on a trip signal. The [7 day] Frequency j ensures that the CEAs are OPERABLE prior to reducing SDM to , less than the limits of LCO 3.1.1. 1 REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50.59.

3. Regulatory Guide 1.68, Revision 2, August 1978.

i l 4. ANSI /ANS-19..1-1985, December 13, 1985.

5. CESSAP, 0" , Chapter 14.

6. 10 CFR 50.46.

7. -CESSAP,O'mChapter 15. m r

t

      % 2 % atetente!- 16A Teek Spec Basee                                                Page B.1.1.E

System 80+ Desian Control Document STE - MODES 1 and 2 B 3.1.10 BASES (continued) ' REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50.59.

3. Regulatory Guide 1.68, Revision 2, August 1978.

4. ANSI /ANS-19.6.1-1985, December 13, 1985.

5. TC';';AR A chapter 14. ,

                                                                                               ~         -

6. CESSAR-0 4 Chapter 15. -m

7. 10 CFR 50.46.

    !a- 'Deelyn 20etenW.16A Tech Spec Resee pay ,y 3,p.gg I

r

System 80+ oestan controlDocument Special Test Exceptions - CEDMS Testing B 3.1.11 l B 3.1.1 REACTIVITY CONTROL SYSTEMS ' B 3.1.11 Special Test Exceptions - CEDMS Testing - - i BASES BACKGROUND Control Element Drive Mechanism System (CEDMS) testing .is performed to verify the operability of the control element drives. Since this test requires the withdrawal of CEAs, -- the SHUTDOWN MARGIN is reduced. In order that the test may be performed, this special test exception is provided. since the requirements of LCO 3.1.1 would be too restrictive to allow performance of the test. l (--n e \ y - -- l F Ref.1, -the conditions of the Control Element Drive > APPLICABLE - Mechanism System (CEDMS) testing were analyzed. It was 9 l SAFETY ANALYSES found that sufficient subcriticality is maintained in cast l of a CEA ejection accident. This is from the fact that - l l prior to testing K(n-1) must be less than 0.99. The margin i j will preclude inadvertent criticality. ! i l LCO Suspension of the SHUTDOWN MARGIN requirement of LCO 3.1.1 may be suspended for pre-startup testing of the CEDMS if l four conditions are met. First, only one CEA may be withdrawn at a time. Second, no CEA may be withdrawn more ! than seven inches. Third, with RTCBs open, K(n-1) must be ; less than 0.99 before the start of testing. Fourth, all ' other operations which involve a reactivity increase must be l suspended during testing. l APPLICABILITY LCO 3.1.11 is applicable during MODES 4 and 5 since these are the MODES during which CEDMS testing is performed. ACTIONS /L1 If any of the four requirements are not met then testing must be suspended and the SHUTDOWN MARGIN must be restored to the limit of LC0 3.1.1. This action is necessary for the prevention of an inadvertent criticality. (continued)

 ?; ADenfpn Atetentet 16A Tech Spec Baser                                         Page B 3.1 62

System 80+ De-ion controlDocument Special Test Exceptions ,CEDMS Testing ], B 3.1.11 , BASES (continued) 1 SURVEILLANCE SR 3.1.11.1 REQUIREMENTS Determination of the SHUTDOWN MARGIN ensures that CEDMS testing is being performed under conditions that would ,a prevent an inadvertent criticality. The Frequency of 24 hours is based upon operating experience and the fact that other administrative controls exist to prevent unauthorized 4 reactivity increases. " i f REFERENCES 1. Safety Evaluation by the Office of NRR, Docket no. STN

;                                   50-530, January 26, 1988.

2. -CE;;A; ;;. Section 19.8, " Shutdown Risk Assessment".

J l i l WM Den &n Matenief- 16A Tech Spec Bases p,,, y 3, y,g4

, System 80+ De' inn controlDocument Boron Dilution Alarms B 3.1.12 BASES

                                                                                                           ~  ~

IACTIONS B.I. B.2. and B.3 i (continued) l With two startup channel high neutron flux alarms inoperable, action must be immediately initiated to restore a single channel to OPERABLE status. Also, the RCS boron concentration must be determined when entering MODE 3, 4, 5, i l 6 or at the time of alarm is determined inoperable. This ' second action is to be performed immediately and once per -- the Frequency _given in the LC0 Tables 3.1.12-1 through 3.1.12-5. Immediate suspension of all operations involving CORE ALTERATIONS or positive reactivity changes is also required. These actions will help prevent the loss of SHUTDOWN MARGIN and return to criticality should an inadvertent boron dilution occur. SURVEILLANCE SR 3.1.12.1 i A CHANNEL CHECK shall be performed on each startup chtnnel once per 12 hours and when initially setting sepoints. This ensures proper operation of the alarms. The Frequency is i based upon operating experience and administrative controls. l ! SR 3.1.12.2 l A CHANNEL CALIBRATION shall be performed on each startup i channel every 31 days of cumult'.ive operation during l shutdown. The Frequency is based upon operating experience. ' l b!EFERENCES 1. --{f5aan9 Chapter 19. 1 l WD=6n a***=w 1sa run s e u sua p,,, a a. s.ee i

   . Syntam 80+                                                          oesten contrat Document                         ~

LF B 3.2. BRSM 4 -- (Bases G. A @ v BACKGROUND Power distribution is a product of multiple parameters, MN - (continued) various combinations of which may produce acceptable power distributions. Operation within the design limits of power M distribution is accomplished by generating operating limits on the LHR and departure from nucleate boiling (DNB). Proximity to the DNB condition is expressed by the departure from nucleate boiling ratio (DNBR), defined as the ratio of ~ the cladding surface heat flux required to cause DNB to the actual cladding surface heat flux. The minimum DNBR value during both normal operation and A00s is [1.24] as calculated by the CE-1 Correlation (Ref. 3) and corrected l for such factors as rod bow and grid spacers. It is accepted as an appropriate margin to DNB for all operating conditions. There are two systems that monitor core power distribution ; online: the Core Operating Limit Supervisory System (COLSS) ' and the core protection calculators (CPCs). The COLSS and CPCs that monitor the core power distribution are capable of verifying that the LHR and the DNBR do not exceed their limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating core power operating limits corresponding to the allowable peak LHR and DNBR. The CPCs perform this function by continuously calculating an actual value of DNBR and local power density (LPD) for comparison with the respective trip j setpoints. A DNBR penalty factor is included in both the COLSS and CPC

DNBR calculations to accommodate the effects of rod bow.

i The amount of rod bow in each assembly is dependent upon the l' average burnup experienced by that assembly. Fuel assemblies that incur higher than average burnup experience j a greater magnitude of rod bow. Conversely, fuel assemblies j that receive lower than average burnup experience less rod bow. In design calculations' for a reload core, each batch of fuel is assigned a penalty applied to the maximum integrated planar radial power peak of the batch. This penalty is correlated with the amount of rod bow determined from the maximum average assembly burnup of the batch. A

;                            single net penalty for the COLSS and CPCs is then determined 4

from the penalties associated with each batch that comprises a core reload, accounting for the offsetting margins due to , the lower radial power peaks in the higher burnup batches. l

(continued)

  ?     .-d Denket Atetend.16A Tech Spec Bases                                             rege g 3.2 2 4                                                                                           . . .            _         --

System 80+ oeskn control Docun ent , 1 LHR B 3.2.1

i l SURVEILLANCE SR 3.2.1.2 (continued)

   'GkJritinebh
   '                           The 31 day Frequency for performance of this SR is consistent with the historical testing frequency of reactor protection and monitoring systems. The Surveillance Frequency for testing protection systems was extended to 92 days by CEN 327. Monitoring systems were not addressed in CEN 327; therefore, this Frequency remains at 31 days.
                                                                                                   ~

REFERENCES 1. CE5 C L Chapter 15. t c

2. 6i55A % Chapter 6. g>., /d

3. CE-1 Correlation for DNBR. "6 E C 548 -

( 4. 10 CFR 50.46, Appendix A, GDC 10. . l l 5. 10 CFR 50.46. i 6. 10 CFR 50.46, Appendix A, GDC 26 l l l l l l

I Anorowd Deskrs Meterial- 16A Tech Spec Bases pay, g 3,24

l l votem 80+ oestan control Document ' F l B3.2.7 l l3.2 POWER DISTRIBUTION LIMITS Fq 3.2.2 Planar Radial Peaking Factors (F,y) , kES $KGR00ND The purpose of this LCO is to limit the core power distribution to the initial values assumed in the accident analyses. Operation within the limits imposed by this LCO either limits or prevents potential fuel cladding failures - that could breach the primary fission product barrier and -< release fission products to the reactor coolant in the event of a loss of coolant accident (LOCA), . loss of. flow accident, ejected control element assembly (CEA) accident, or other postulated accidents requiring termination by a Reactor . Protection System (RPS) trip function. This LCO limits damage to the fuel cladding during an accident by ensuring i that the plant is operating within acceptable conditions at ' the onset of a transient. Methods of controlling the power distribution include: I a. Using full or part strength CEAs to alter the axial power distribution;

b. Decreasing CEA insertion by boration, thereby improving the radial power distribution; and l c. Correcting off optimum conditions (e.g., a CEA drop or l l' misoperation of the unit) that cause margin !

degradations. , The core power distribution is controlled so that, in l conjunction with other core operating parameters (CEA insertion and alignment limits), the power distribution does not result in violation of this LCO. Limiting safety system settings and this LCO are based on the accident analyses (Refs. I and 2), so that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences (A00s), and the limits of acceptable consequences are not exceeded for other postulated accidents. (continued) r l M W Alesenfel- 16A Tech Spec Benee ene ge g 2.2.g

Syntem 80+ oesinn controlDocument F B 3.2.7 l Bases (continued) SURVEILLANCE 1R 3.2.2.1 -l REQUIREMENTS l This periodic Surveillance is for determining, using the Incore Detector System, that Fly values are s Fly values used in the COLSS and CPCs. It ensures that the Fly values used remain valid throughout the fuel cycle. A Frequency of ; 31 EFPD is acceptable because the power distribution changes J only slightly wit 5 the amount of fuel burnup. Determining . 1 the Ffy values after each fuel loading when THERMAL POWER is -

                              > 40% RTP, but prior to its exceeding 70% RTP, ensures that the core is properly loaded.
                                                                                                     ==

REFERENCES 1. 6 Chapter 15.

2. M Chapter 6.

3. CE-1 Correlation for DNBR. i

4. 10 CFR 50.46, Appendix A, GDC 10.

5. 10 CFR 50.46.

6. 10 CFR 50, Appendix A, GDC 26.

i l AnwoM Design Materiel- 16A Tech Spec Bases page g 3.2.g5

System 80+ Deslan ControlDocument l T B3.2.3 D gases b + A1) - BACKGROUND Power distribution is a product of multiple parameters, '; (continued) various combinations of which may produce acceptable power 4[g , distributions. Operation within the design limits of power l distribution is accomplished by generating operating limits i on the linear heat rate (LHR) and the departure from j nucleate boiling (DNB). i Proximity to the DNB condition is expressed by the departure " from nucleate boiling ratio (D.':BR), defined as the ratio of the cladding surface heat flux required to cause DNB to the actual cladding surface heat flux. The minimum DNBR value , during both normal operation and A00s is [1.24] as- ' calculated by the CE-1 Correlation (Ref. 3) and corrected for such factors as rod bow and grid spacers, and it is accepted as an appropriate margin to DNB for all operating conditions. There are two systems that monitor core power distribution - online: the Core Operating Limit Supervisory System (COLSS) and the core protection calculators (CPCs). The COLSS and CPCs that monitor the core power distribution are capable of verifying that the LHR and the DNBR do not exceed their limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating core power operating limits corresponding to the allowable peak l LHR and DNBR. The CPCs perform this function by continuously calculating actual values of DNBR and local , power density (LPD) for comparison with the respective trip i setpoints.

l A DNBR penalty factor is included in the COLSS and CPC DNBR ) l i calculation to accomraodate the effects of rod bow. The amount of rod bow in each assembly is dependent upon the average burnup experienced by the assembly. Fuel assemblies that incur higher than average burnup experience greater magnitude of rod bow. Conversely, fuel assemblies that receive lower than average burnup experience less rod bow. In design calculations for a reload core, each batch of fuel is assigned a penalty applied to the maximum integrated planar radial power peak of the batch. This penalty is correlated with the amount of rod bow that is determined from the maximum average assembly burnup of the batch. A single net penalty for the COLSS and CPCs is then determined from the penalties associated with each batch that comprises (continued) L-. : . : Denkn Atetodel 16A Tech Spec Bases pay, g 3,2.g7 l

System 80+ oeskn controloccument B3.2.$ as APPLICABLE. Fuel cladding damage does not occur from conditions outside ' SAFETY ANALYSES the limits of these LCOs during normal operation. However, (continued) fuel cladding damage could result if an accident occurs due to initial conditions outside the limits of these LCOs. Th potential for fuel cladding damage exists because changes i the power distribution can cause increased power peaking an. correspondingly increased local LHRs. T, satisfies Criterion 2 of the NRC Policy Statement. LCO The power distribution LCO limits are based on enrrelations between power peaking and certain measured varialles used as inputs to the LHR and DNBR operating limits. The power distribution LC0 limits are provided in the COLR. The limitations on the T are provided to ensure that design ~ operatingmarginsaremalntained. T > 0.10 is not expected. If it occurs, the actions,to be taken ensure that operation is restricted to only those conditions required to identify the cause of the tilt. It is necessary to explicitly account ( " power asymmetries because the radial peaking factors usec in the core power distribution calculations are based on an untilted power distribution. APPLICABILITY Power distribution is a concern any time the reactor is critical. The power distributien LCOs, however, are only applicable in MODE 1 above 20% RTP. The reasons these LCOs are not applicable below 20% RTP are:

a. The incore neutron detectors that provide input to the COLSS, which then calculates the operating limits, are inaccurate due to the poor signal to noise ratio that -

they experience at relatively low core power levels.

b. As a result of. this inaccuracy, the CPCs assume a minimum core power of 20% RTP when generating LPD and DNBR trip signals. When the core power is below this !

level, the core is operating well below its thermal ' limits and the resultant CPC calculated LPD and DNBP. trips are highly conservative. (continued) WDwign AcetwW 16A Tuh Spu Buu p,,, y 3,g ,y

I Sy2 tem 80+ Deelms CorrtrolDvcumerrt l i [ B 3.2 I D e / 4 /e I Ba s # N i u & ,.

l ACTIONS i

B.I. B.2. and B.3 (continued) ) If the measured T is not restored to within its specified

limits, the react,or continues to operate with an axial power i distribution mismatch. Continued operation in this j configuration may induce an axial xenon oscillation, which J results in increased linear heat generation rates when the

! xenon redistributes. If the measured T cannot be restored to within its limit withia 2 hours, realtor power must be I I reduced. Reducing THERMAL POWER to < 50% RTP within 4 hours ! provides an acceptable level of protection from increased , I power peaking due to potential xenon redistribution while 1 maintaining a power level sufficiently high enough to allow l 4 the tilt to be analyzed. 1 The Linear Power Level-High trip setpoints are reduced to 1 s 55% RTP to ensure that the assumptions of the accident analysis regarding power peaking are maintained. After i

power has been reduced to s 50% RTP, the rate and magnitude of changes in the core flux are greatly reduced. Therefore, i j 16 hours is an acceptable time period to allow for reduction j of the Linear Power Level-High trip setpoints, Required s Action B.2. The 16 hour Completion Time allowed to reduce i the Linear Power Level-High trip setpoints is required to

l perform the actions necessary to reset the trip setpoints.

i i ! THERMAL POWER is restricted to ET RTP until the measured T' l is restored to within its specified limit by correcting the ! ! out of limit condition. This action prevents the operator ! j from increasing THERMAL POWER above the conservative limit i ! when a significant T has existed, but allows the unit to i continueoperationf8rdiagnosticpurposes. l J l The Completion Tint of Required Action B.3 is modified by a l Note governing subsequent power increases. After a THERMAL f POWER increase foll> wing restoration of T , operation may i proceed provided t:ie measured T is determined to remain I withinitsspecifsedlimitattheincreasedTHERMALPOWER level. i l l [ (continued) 4

wnDee4pn neneerw. tan recs spec sneen % s 2.2-22

i System 80+ Deslan controlDocument B3.2.$ Bases ) SURVEILLANCE SR 3.2.3.2 (continued) -

                                                                                                                         -l REQUIREMENTS protection and monitoring syrtems. The Surveillance Frequency for testing protection systems was extended to 92 days by CEN 327. Monitoring systems were not addressed in CEN 327; therefore, this Frequency re. mains at 31 days.

I 3.2.3.3 SR Independent confirmation of the validity of the COLSS : calculated T, ensures that the COLSS accurately identifies l T,' s . The 31 day Frequency for performance of this SR is consistent with the historical testing frequency of reactor protection and monitoring systems. The Surveillance Frequency for testing protection systems was extended to 92 days by CEN 327. Monitoring systems were not addressed in CEN 327; therefore, this Frequency remains at 31 days. REFERENCES. 1. -- CESSAR- % Chapter 15. *

2. TETJ"" DC Chapter 6. 4 -

3. CE-1 Correlation for DNBR.

4. 10 CFR 50.46, Appendix A, GDC 10.

5. 10 CFR 50.46.

6. 10 CFR 50, Appendix A, GDC 26.

 \pproved Destpro Mstenle! 16A Tech Spec Basee 4 y gg.y

- Syntem 80 + Deslan controlDocument DNBR

  ,- -          ,                                                                            B 3.2.4                                 i n>

ase y u{q,y,, g, g ,

,ACKGROUND Limiting power distribution skewing over time also minimizes * (continued) the xenon distribution skewing, which is a significant factor in controlling axial power distribution. Power distribution is a product of multiple parameters, , various combinations of which may produce acceptable power ; distributions. Operation within the design limits of power i distribution is accomplished by generating operating limits ~~ - on the linear heat rate (LHR) and the departure from ; nucleate boiling (DNB). j Proximity to the DNB condition is expressed by the DNBR, defined as the rktio of the cladding surface heat flux required to cause DNB to the actual cladding surface heat l flux. i The minimum DNBR value during both normal operation and A00s is [1.24] as calculated by the CE-1 Correlation (Ref. 3) and - l corrected for such factors as rod bows and grid spacers and ; it is accepted as an appropriate margin to DNB for all operating conditions. There are two systems that monitor core power distribution ! online: the Core Operating Limits Supervisory System (COLSS) and the core protection calculators (CPCs). The COLSS and CPCs that monitor the core power distribution are capable of verifying that the LHR and DNBR do not exceed ! their limits. The COLSS performs this function by ! continuously monitoring the core power distribution and calculating core power operating limits corresponding to the { allowable peak LHR and DNBR. The CPCs perform this function ; by continuously calculating an actual value of DNBR and LPD ' for comparison with the respective trip setpoints. A DNBR penalty factor is included in both the COLSS and CPC DNBR calculation to accommodate the effects of rod bow. The ; amount of rod bow in each assembly is dependent upon the ' average burnup experienced by that assembly. Fuel assemblies that incur higher than average burnup experience a greater magnitude of rod bow. Conversely, fuel assemblies that receive lower than average burnup experience less rod bow. In design calculations for a reload core, each batch of fuel is assigned a penalty that is applied to the maximum integrated planar radial power peak of the batch. This penalty is correlated with the amount of rod bow that is (continued) noemt onow aneeerw. ran rees spee sneen rene s 2.2 2s

System 80+ Deslan Control Document DNBR B 3.2.4 Bases'(continued) _ REFERENCES 1. - CE55AR % Chapter 15. '

2. CE';';AR 'L Chapter 6.

3. CE-1 Correlation for DNBR.

4. 10 CFR 50, Appendix A, GDC 10.

5. 10 CFR 50.46.

6. 10 CFR 50, Appendix A, GDC 26.

9 l l eh e 9 4pwoM Dukn Mahwiel 16A Tech Spec Bues pay,y3,3.h33

System 80+ De%n CrntrolDocument ASI B 3.2.5 r ases j p Lk % M l e * ' 4 's ) BACKGROUND Power distribution is a product of multiple parameters, l (continued) various combinations of which may produce acceptable power ) distributions. Operation within the design limits of power l distribution is accomplished by generating operating limits on the linear heat rate (LHR) and the departure from nucleate boiling (DNB). I Proximity to the DNB condition is expressed by the departure ! from nucleate boiling ratio (DNBF p, defined as the ratio of -t' the cladding surface heat flux required to cause DNB to the actual cladding surface heat flux. The minimum DNBR value , during both normal operation and A00s is [1.24] as calculated by the CE-1 Correlation (Ref. 3), and corrected for such factors as rod bow and grid spacers, and it is accepted as an appropriate margin to DNB for all operating conditions. [ There are two systems that monitor core power distribution i online: the Core Operating Limit Supervisory System (COLSS) ; or the core protection calculators (CPCs). The COLSS and l CPCs monitor the core power distribution and are capable of l verifying that the LHR and DNBR do not exceed their limits. 2 The COLSS performs this function by continuously monitoring , the core power distribution and calculating core power ' operating limits corresponding to the allowable peak LHR and DNBR. The CPCs perform this function by continuously calculating actual values of DNBR and local power density (LPD) for comparison with the respective trip setpoints. A DNBR penalty factor is included in both the COLSS and CPC < DNBR calculations to accommodate the effects of rod bow. The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly. Fuel assemblies that incur higher than average burnup experience greater rod bow. Conversely, fuel assemblies that receive lower than average burnup experience less rod bow. In i design calculations for a reload corc, each batch of fuel is assigned a penalty that is applied to the maximum integrated planar radial power peak of the batch. This penalty is correlated with the amount of rod bow that is r / orzined from the maximum average assembly burnup of th atch. A single net penalty for the COLSS and CPC is tb 9 Jetermined from the penalties associated with each batch t..at comprises a core reload, accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches. (continued) Anewoved Design nieseniel.16A Tech Spec Benee

                                                                                        % B 3.2 36

System 80+ Deslan connot Document

, ASI l B 3.2.5 i jBases(continued) 1 jREFERENCES 1. mEssany Chapter 15. I 2. CESSAR- R Chapter 6. i 3. CE-1 Correlation for DNBR. ; I ;

4. 10 CFR 50, Appendix A, GDC 10. I

5. 10 CFR 50.46.

6. 10 CFR 50, Appendix A, GDC 26.

l 2 i e 1 a W Dee5, A0esenfel- 16A Tech Spec Bosee pp y 194

9 i System 80+ Deslan ControlDocument RPS Instrumentation - Operating B 3.3.1 , BASES ;

                                       ~
                                                                                                                                                       ~      ~
     ~
                  ' APPLICABLE                        ~ 6, 7. Steam Generator Pressure - Low (continued)                                                                ,

SAFETY ANALYSES i

                        %                                         needed to shut down the reactor and assist the ESF System in the event of an HSLB or' main feedwater 1ine. break- accihnt<                           . . ~ - . . ~ . . ~                .--...--            ~.

i 8, 9. Steam Generator Level - Low l D-el d ~ The Steam Generator #1 Level - Low and Steam Generator #2 Level - Low trips ensure that a reactor - trip signal is generated for the following events to , help prevent exceeding the design pressure of the bhM RCS due to the loss of the heat sink: -

Lo,ss of Normal,Feedwater Event (A00); and.

         -4                                  .j.                  *      ;- Feedwater System, Pipe Break (Accident).                                                 ,

{ 10, 11. Steam Generator Level - Hiah -

                                                                                                                                                                              }
                                                                                                        .,..                        m

.= The.Stcan

                                                                     ~ . ~

GenernorJ1 L'evel - High and4 Steam: .s . Generator'J2 Level *~ 'High trips are pr'o vided to protect the turbine'from excessive moisture- t carryover in case -of a~ steam generator'. overfill event.

12. Reactor Coolant Flow - Low The Reactor Coolant Flow - Low trip provides protection against an RCP Sheared Shaft Event. The DNBR limit may be exceeded during this event; however, the trip ensures the consequence.s are acceptable.

                                                                                                                                       .c                          .

13. Local Power Density - Hiah' ,

The CPCs perform the chlculations required to derive the DNBR and LPD parameters, and their associated-RPS trips. The DNBR - Low and LPD - High trips provide plant protection during the following A00s and assist the ESF systems in the mitigation of the folicwing accidents. (continued) Approves nesnst Meterl+1 16A Tre'

nee Pope g 3,3 12

                                                                                                                     - , , -        _<                               j      ,

l Syntem 80+ Deslan ControlDocument t RPS Instrumentation - Operating 4 8 3.3.1 i BASES ne j APPLICABLE 13. Local Power Density - Hiah (continued) y SAFETY ANALYSES d l i W The LPD - High trip provides protection against fue' centerline melting due to the occurrence of excessive local power density peaks during the sa Q following A00s: pg

Decrease in Feedwater Temperature; _
Increase in Feedwater Flow; Increased Main Steam Flow (not due to the steam line rupture) Without Turbine Trip;
Uncontrolled CEA Withdrawal From Low Power;
Uncontrolled CEA Withdrawal at Power; ,
CEA Misoperation; Single Part Strength CEA Drop;
CEA Misoperation; Full-strength and part-strength CEA subgroup drop; and
CEA Misoperation; Out-of-sequence operation.

For the events listed above (except CEA Misoperation; Single Part Strength CEA Drop) DNBR4ow will trip the reactor first, since DNB - would occur before fuel centerline melting would occur.

14. Deoarture from Nucleate Boilina Ratio (DNBR) - Low '

The CPCs perform the calculations required to derive the DNBR and LPD parameters, and their associated - RPS trips. The DNBR - Low and LPD - High trips ; provide plant protection during the following A00s and assist the ESF systems in the mitigation of the following accidents. l l I (continued) W Deekt Atesud.16A Tech Spec Seeen page B 2.2 1g l

I h

System 80+ D& Control Document RPS Instrumentation - Operating l i B 3.3.1 l BASES *X i ,

14. Departure from Nucleate Boilina Ratio (DNBR) - Low '

l; SAFETY APPLICABLE ANALYSES (continued) N lM

Uncontrolled boron dilution; 1
Out-of-Sequence insertion or withdrawal of CE/

l group. h - a Interlocks /Bvoasses The bypasses and their Allcwable Values are addressed in footnotes to Table 3.3.1-1. They are not otherwise addressed as specific Table entries. The automatic operating bypass removal features must function as a backup to manual actions for all safety related trips to ensure the trip Functions are not operationally bypassed when the safety analysis assumes the

Functions are not bypassed. The basis for each of the operating bypasses is discussed under individual trips in the LCO section:

a. Logarithmic Power Level - High; l

b. DNBR - Low and LPD - High; and

c. Pressurizer Pressure - Low.

The RPS satisfies Criterion 3 of the NRC Policy Statement. ,LC0 The LCO requires all instrumentation performing an RPS Function to be OPERABLE. Failure of any required portion of the instrument CHANNEL renders the affected CHANNEL (s) l inoperable and reduces the reliability of the affected Functions. t Actions allow maintenance (TRIP CHANNEL) bypass of individual CHANNELS but the bypass activates interlocks that prevent operation with a second CHANNEL in the same Function bypassed. With one CHANNEL in each Function TRIP CHANNEL bypassed, this effectively places the plant in a two-out-of-three logic configuration in those Functions. (continued) anome one6n unennw. tu rece spec sua rene a 2.2-si

System 80+ Deslan ControlDocument i

RPS Instrumentation - Operating B 3.3 iBASES i hM

M -

LCO 4. Pressurizer Pressure - Low (continued) i The Allowable Value is set low enough to prevent a i reactor trip during normal plant operation and pressurizer pressure transients. However, the S" M setpoint is high enough that with a LOCA, the F M -e-reactor trip will occur soon enough to allow the ESF systems to perform as expected in the analyses . _ and mitigate the consequences of the accident.

The trip setpoint may be manually decreased to a
minimum value (floor value) of [300 psia) as pressurizer pressure is reduced during controlled plant shutdowns, provided the margin between the pressurizer pressure and the setpoint is maintained less than [400 psia). This allows for controlled depressurization of the RCS while still maintaining an active trip setpoint until the time is reached when the trip is no longer needed to protect the plant. Since the same Pressurizer Pressure - Low bistable is also shared with the SIAS, an inadvertent SIAS actuation is also prevented. The setpoint increases automatically as pressurizer pressure increases, until the trip setpoint is reached.

The Pressurizer Pressure - Low trip and the SIAS Function may be simultaneously bypassed when RCS pressure is below [400 psia], when neither the reactor trip nor an inadvertent SIAS actuation are desirable, and these Functions are no longer needed to protect the plant. The bypass is automatically removed as RCS pressure increases above [500 psia). The difference between the operating bypass enable and removal features allows for bypass permissive bistable hysteresis, and allows setting the operating bypass setpoint close enough to the limit so as to avoid inadvertent actuation at the [300 psia] trip setpoint minimum value (floor value).

5. Containment Pressure - Hiah The LCO requires four CHANNELS of Containment pressure-High to be OPERABLE in MODES 1 and 2.

(continued) nemmtw neww ssa ruh seu sua reo, a 2.2 21

Srtem 80+ Deslan ControlDocument RPS Instrumentation - Operating B 3.3.1 BASES b m - LCO 14. Departure from Nucleate Boilina Ratio (DNBR)-Low g (continued) This operating bypass is required to perform a plant startup, since both CPC generated trips wil' be in effect whenever shutdown CEAs are inserted. It also allows system tests at low power with Pressurizer Pressure - Low or RCPs off. _ During special testing pursuant to LCO 3.4.17, the CPC CHANNELS may be manually bypassed when THERMAL POWER is below [5%) RTP to allow special testing without generating a reactor trip. The Variable Overpower - High trip setpoint is reduced, so as to provide protection during testing. Interlocks / Bypasses The LCO on operating bypass permissive removal CHANNELS requires that the automatic operating bypass removal feature ' of all four operating bypass CHANNELS be OPERABLE for each RPS Function with an operating bypass in the MODES addressed in the specific LCO for each Function. All four operating bypass removal CHANNELS must be OPERABLE to ensure that none of the four RPS CHANNELS are inadvertently bypassed. This LCO applies to the operating bypass. removal feature only. If the bypass enable Function is failed so as to prevent entering a bypass condition, operation may continue. In the case of the Logarithmic Power Level - High trip (Function 2), the absence of a bypass will limit maximum power to below the trip setpoint. The interlock function Allowable Values are based upon analysis of functional requirements for the bypassed Functions. These are discussed above as part of the LCO discussion for the affected Functions. APPLICABILITY The Variable Overpower - High, Pressurizer Pressure - High, Pressurizer Pressure - Low, Containment Pressure - High, Steam Generator Level - Low, and Steam Generator Level - High trips are required to be OPERABLE in MODES 1 and 2 because the reactor is critical in these MODES. The reactor (continued) w a ono6ru neeuw. saa rees soec sua noea2.221

' Syntem 80 + Design ControlDocument RPS Instrumentation - Operati B 3.3 BASES D0( M - , SURVEILLANCE RPS Initiation Loaic Testina (continued) REQUIREMENTS *W M The quarterly CPC CHANNEL FUNCTIONAL TEST is performed us-software. This software includes preassigned addressable constant values that may differ from the current values. Provisions are made to store the addressable constant values on a computer disk prior to testing and to reload them after testing. A Note is added to the Surveillance Requirements _. to verify that the CPC CHANNEL FUNCTIONAL TEST includes the correct values of addressable constants. The 92 day surveillance interval is based upon the experience with safety related computer systems for operating plants. SR 3.3.1,9 A Note indicates that neutron detectors are excluded from CHANNEL CALIBRATION. A CHANNEL CALIBRATION of the power range neutron flux channel every 92 days ensures that the channels are reading accurately and within tolerance. The Surveillance verifies that the channel responds to a measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account , for instrument drifts between successive calibrations to j ensure that the channel remains operational between ; successive tests. Measurement error determination, setpoht l error determination, and calibration adjustment must be i performed consistent with the plant specific setpoint analysis. The channel shall be left calibrated consistent l with the assumptions of the current plant specific setpoint analysis. The as found and as left values must be recorded and reviewed for consistency with the assumptions of the interval between surveillance interval analysis. The requirements for this review are outlined in Reference [9]. The Frequency is based upon the assumption of an (18] month calibration interval for the determination of the magnitude of equipment drift in the setpoint analysis as well as operating experience and consistency with the typical (18] month fuel cycle. The detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a (continued) toprowd Design Material- 16A Tech Spec Bases Page B 3.3-29

STtem 80 + Deslan CwntrolDocument . RPS Instrumentation - Operating I l B 3.3.1 ; i BASES l SURVEILLANCE SR 3.3.1.15 *l l l REQUIREMENTS (continued) This SR ensures that the RPS RESPONSE TIMES are verified to , l be less than or equal to the maximum values assumed in the l l safety analysis. Individual component response times are i not modeled in the analyses. The analyses model the overall l or total elapsed time, from the point at which the parameter l exceeds the trip setpoint value at the sensor, to the point _. t at which the RTCBs open. Response times are conducted on an [18] month STAGGERED TEST BASIS. This results in the interval between successive surveillances of a given CHANNEL of n x 18 months, where n is the number of CHANNELS in the i function. The Frequency of 18 months is required because : response times cannot be determined at power, since equipment operation is required. Testing may be performed in one measurement or in overlapping segments, with 1 verification that all components are tested. 1 A Note is adoed to indicate that the neutron detectors may l be excluded from RPS RESPONSE TIME testing because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal. Slow changes in detector sensitivity are compensated for by performing the daily calorimetric calibration (SR 3.3.1.5). ' REFERENCES 1. 10 CFR 50, Appendix A, GDC 21.

2. 10 CFR 100.

3. Removed

4. IEEE Standard 279-1971, April 5,1972.

5. WESjlR=Mg S Chapter 15.

I

6. 10 CFR 50.49.

7. [Setpoint Report).

8. M Chapter 7. z .

(continued) Approwd DeeJpn Atatoriel.16A Tech Spec Bases Pope B 3.343 l

  . System 80+                                                      Deslan ControlDocument RPS Instrumentation - Shutdown B 3.3.2 BASES APPLICABLE          1.      Loaarithmic Power level - Hiah (continued)

SAFETY ANALYSES g dQ RCPB in the event of an unplanned criticality from a 3M% shutdown condition. g In MODES 2, 3, 4, and 5, with the RTCBs closed, and the Control Element Assembly (CEA) Drive System .. capable of CEA withdrawal, protection is required for APPL Ir **LE . CEA withdrawal events originating when THERMAL POWER g, - , 4 % ,a is < [lE-4%] RTP. For events originating above this power level, other trips provide adequate protection. ""fi s.4-s7 MODES 3, 4, and 5, with the RTCBs closed, are addressed in this LCO. MODE 2 is addressed in : I LCO 3.3.1. In MODES 3, 4, or 5, with the RTCBs open or the CEAs not capable of withdrawal, the Logarithmic Power Level

                             - High trip does not have to be OPERABLE. However, the iridication and alarm portion of two logarithmic                                                      l channels must be OPERABLE to ensure proper indication of neutron population and to indicate a boron dilution event. The indication and alarm functions are                                                             l addressed in LC0 3.3.13.

2,3. Steam Generator Pressure - Low The Steam Generator #1 Pressure - Low and Steam Generator #2 Pressure - Low trips provide protection against an excessive rate of heat extraction from the steam generators and rcsulting rapid, uncontrolled cooldown of the RCS. This trip is needed to maintain shutdown conditions and assist the ESF System in the event of an MSLB while shutdown.

4. , Reactor Coolant Flow - Low The Reactor Coolant Flow-Low trip provides protection against Excess Heat Removal Events while shutdown.

This trip will ensure that the plant is in a

configuration (i.e., at least one RCP running in each loop) which results in acceptable consequences of an Excess Heat Removal Event.

(continued) no wemtouen www- tsA rocs spec sun rene a 2.2-ss

  -       . .       .- .-        _~     _

l l Syntem 80+ Deslan ControlDocument I RPS Instrumentation j B 3.3.2 BASES l j j ACTIONS A.1 and A.2 (continued) i 2 The Completion Time of prior to entering MODE 2 following i the next MODE 5 entry is based on adequate channel to channel independence, which allows operation with two or i more channels since no single failure will prevent a reactor

trip.

j __ u l Condition B applies to the failure of two channels in any

RPS automatic trip Function.

J 1 The Required Action is modified by a Note stating that LCO 3.0.4 is not applicable. The Note was added to allow

the changing of MODES, even though two channels are inoperable, with one channel bypassed and one tripped. In: this configuration, the protection system is in a one-out-J of-two logic, which is adequate to ensure that no random failure will prevent protection system operation.

l ) h Required Action B.1 provides for placing one inoperable i channel in bypass and the other channel in trip within the j Completion Time of 1 hour. This Completion Time is j sufficient to allow the operator to take all appropriate i actions for the failed channels while ensuring the risk j involved in operating with the failed channels is

acceptable. With one channel of protective instrumentation t bypassed, the RPS is in a two-out-of-three logic; but with j another channel failed, the RPS may be operating in a two-out-of-two logic. This is outside the assumptions made in i the analyses and should be corrected. To correct the i

problem, the second channel is placed in trip. This places ! the RPS in a one-out-of-two logic. If any of the other

OPERABLE channels receives a trip signal, the reactor will i trip.

l One of the two inoperable channels will need to be restored to operable status prior to the next required CHANNEL FUNCTIONAL TEST, because channel surveillance testing on an l OPERABLE channel requires that the OPERABLE channel be i placed in bypass. However, it is not possible to bypass more than one RPS channel, and placing a second channel in g~A___,k j trip will result in a reactor trip. Therefore, if one RPS M 4 -8. ~4 C ~ k (continued) I" g V

  ?n.2 % A000eriel* 16A Tech Spec Bases                                           hago G 3.3 66 wer

 . ~ -       . -     -       - . - - . . .   .     -       -        ..      --      . - .                 - . --

l System 80+ 0+n ControlDocument RPS Instrumentation . B 3.3.2 O . / (c.o d A IASES d'

                                                     'K                                                ~T[,a p            -

C.1. C.2.1. an # p 8. I rm SCTIONS *l lm - -

channel is in trip and a second channel is in bypass, a 1

4 third inoperable channel would place the unit in LCO 3.0.3. m // i T + Condition C applies to one automatic operating bypass removal function inoperable. If the inoperable bypass

                                                                                                     @ S'S-

removal function for any TRIP CHANNEL cannot be restored to gg i bj /. OPERABLE status within I hour, the associated RPS channel hA ,

l may be considered OPERABLE only if the bypass is not in l ~3 effect. The operator must verify that the operating bypass

'g. / 3 is not in effect within one hour and every 12 hours thereafter; otherwise the affected RPS channel must be f declared inoperable, as in Condition A, and the affected automatic TRIP CHANNEL placed in bypass or trip. The l

operating bypass removal function and the automatic TRIP l CHANNEL must be repaired prior to entering MODE 2 following _ the next MODE 5 entry. The Bases for the Required Actions

and Required Completion Times are consistent with Condition A. The Required Action is modified by a Note stating that this LCO applies only to Functions 1, 5, and 6. This Note aids in identifying the applicable functions; Logarithmic Power Level - High, Reactor Coolant Flow - Low, LPD - High, and DNBR - Low. 0.1 and 0.2 Condition D applies to two inoperable automatic operating bypass removal functions. If the operating bypass removal functions for two operating by OPERABLE status within I hour,the passes cannot be restored to associated TRIP CHANNEL may be considered OPERABLE only if the operating bypasses are not in effect. The operator must verify that the operating bypass is not in effect within one hour and every ) 12 hours thereafter; otherwise the affected RPS channels must be declared inoperable, as in Condition B, and the operating bypasses either removed or one automatic TRIP CHANNEL placed in bypass and the other in trip within I hour. The restoration of one affected bypassed automatic trip channel must be completed prior to the next CHANNEL FUNCTIONAL TEST, or the plant must shut down per LC0 3.0.3 as explained in Condition B. (continued) M Denkrr nieserial.16A Tech Spec Reser Pege B 3.3-67

Syntem 80+ Design ControlDocument RPS Instrumentation B 3.3.2 BASES REFERENCES 4. @ Section 7.2. ' (continued)

5. 10 CFR 50.49.

6. [Setpoint Report).

7. CEN-327, June 2, 1986, including Supplement 1, March 3, 1989.

8. [Surveillarc *

                                                        -val Analysis.]

9. Section J.8.

Anwoved Deekn AceterW 16A Tech .%eec Reser Pege B 3.3-75

Sy~ tem 80+ Design ControlDocument CEAl B 3.3 BASES W c_N ACTIONS M (continued) ' flux control. This verification ensures that undesired perturbations in local fuel burnup are prevented. M

The "RSPT/CEAC Inoperable" addressable constant in each of the CPCs is set to indicate that both CEACs are inoperable. This provides a conservative penalty factor to ensure that a conservative effective margin is maintained by the CPCs in the computation of DNBR and LPD trips. M The CEDMCS is placed and maintained in " STANDBY," except during CEA motion permitted by Required Action B.2, to prevent inadvertent motion and possible misalignment of the CEAs. M A comprehensive set of comparison checks on individual CEAs l within groups must be made within 4 hours. Verification l that each CEA is within 7 inches of other CEAs in its group l provides a check that no CEA has deviated from its proper position within the group. M Condition C is entered when the Required Action and associated Completion Time of Condition B is not met. If the Required Actions associated with this Condition cannot be completed within the required Completion Time, the reactor must ba brought to a MODE where the Required Actions do not apply. ;he Completion Time of 6 hours is reasonable, based on operating experience, for reaching the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. (continued) Approved Design Material 16A Tech Spec Bases Page B 3.3-81

l l Syntem 80+ Desian ConholDocument CEACs B 3.3.3 BASES i . . ! SURVEILLANCE SR 3.3.3.6 (continued) REQUIREMENTS are undetectable by the self monitoring function. Operating ; experience has shown that undetected CEAC failures do not ' I occur in any given [18] month interval. REFERENCES 1. 10 CFR 50. 2, 10 CFR 100.

3. % Section 7.2.

4. [ Surveillance Interval Analysis.]

I 4 proved Des.'gre Mowiel 16A Tech Spec Bases rege g 3.3.g5

s System 80+ Denfan ControlDocument ESFAS Instrumentation l B 3.6.5 i BASES BACKGROUND Bistable Processors " * (continued) ' The trip signal is generated by the Bistable Logic processors which compare the input signals to either fixed or variable set points. These bistable outputs for each ; parameter (e.g. Pressurizer Pressure, Steam Generator. Level, etc.) are sent to Local Coincidence Logic where two-out-of-

four logics are perforned. Bistable trip generation is .. described in CESSAR-DC, Section 7.3 (Ref.1). - The trip setpoints and Allowable Values used in the bistables are based on the analytical limits stated in (Ref. 5). The selection of these trip setpoints is such that adequate protection is provided when all sensor and processing time delays are taken into account. To allow for calibration tolerances, instrumentation uncertainties, instrument drift, and severe environment effects, for those-ESFAS channels that must function in harsh environments as defined by 10 CFR 50.49 (Ref. 6), Allowable Values ' , specified in Table 3.3.5-1, in the accompanying LCO, are conservatively adjusted with respect to the analytical ' limits. A detailed example of the methodology used to calculate the trip setpoints, including their explicit uncertainties, is provided in the-[Setpoint Report) (Ref. 7). The actual nominal trip setpoint entered into the bistable is normally still more conservative than that specified by the Allowable Value to account for chances in random measurement errors detectable by a CHANNEL FU'4CTIONAL TEST. A channel is inoperable if its actual trip setpoint is not within its required Allowable Value. Setpoints in accordance with the Allowable Value will ensure that Safety Limits of LCO Section 2.0, " Safety Limits," are not violated during A00s and the consequences of Design i ' Basis Accidents (DBAr.) will be acceptable, providing the plant is operated fr.n within the LCOs at the onset of the r" A00 or DBA and the equipment functions as designed.

                                                                                                          $s       u d

r BPASSk %[ ] 4 The trip channel bypasses and operating bypasses are w manipulated by the Interface and Test Processor in each s R

                                                                                                                  'N channel.

m} @ (continued) L:Dee w ,anoeu w tsa recs spee seene , , ,,,,,,,

System 80+ Design ControlDocument ESFAS Instrumentation 1 B 3.6.5 4 BASES BACKGROUND BMASSE% (continued) L The trip channel bypass p k trip from 4 contributing to the initi g- e action. The trip channel bypass infon q to four channels of Local Coincidence Logi D logic into two-out-of-three by Interface rs. The LCLs only allow one channel byy . . .. .. ._ In addition to the trip channel bypasses, there are also operating bypasses on low pressurizer pressure. These bypasses are enabled manually, in all four PPS channels, when plant conditions do not warrant the specific trip

protection. All operating bypasses are automatically
removed when enabling bypass cor.ditions are no longer satisfied.

The status of any bypass is indicated at the PPS channel

cabinet and the PPS Remote Operator's Module in the control . room. In addition, all operating bypasses and a summary of

the bistable trip channel bypasses in each channel are made available for control room indication via PPS Operator's Module, DIAS and DPS. CESSAR-DC, Section 7.3 (Ref. I) provides a detailed description of these bypasses.

Functional testing of the ESFAS, from the bistable input through the opening of initiation relay contacts in the ESFAS Actuation Logic, can be performed either at power or at shutdown, and is normally performed on a quarterly basis. CESSAR-DC, Section 7.3 (Ref. I) provides more detail on ESFAS testing. Process transmitter calibration is normally performed on a rafueling basis. SRs for the channels are specified in th uirements section. ESFAS thG M 4 .f - The ESFAS Logic D

                                                                   )GIC CHANNEL, ACTUATION L0fi!C, and, C06         W            IIC employs a scheme that provides an ESI            n           trains when bistables in any ?.wo of the                       sing the same input parameter trip            {y          a two out of four trip logic.                    tu (continued)

I Approved Desiers Material 16A Tech Spec Bases page a 3.3 102

i l Syntem 80+ Design ControlDocument ESFAS Instrumentation B 3.6.5 BASES APPLICABLE 5, 6. Emeraency Feedwater Actuation Sianal (Gudh T

SAFETY ANALYSES

   .4sogimtadte                       EFAS maintains a steam generator heat sink during a i                                      loss of MFW event, steam generator tube rupture
           $                          event, MSLB, or FWLB event either inside or outside d                containment, or any event where normal AC power or the MFW system is unavailable. EFAS is also initiated by a loss of power to two or more like              ..

measurement channels. h n# v Low steam generator water level initiates emergency feed to the affected steam generator. If the affr.ed steam generator recovers the level high enoun then the high level signal terminates the emergency feedwater flow to the affected steam generator. l ~ l The ESFAS satisfies Criterion 3 of the NRC Policy

Statement.

LC0 The LCO ensures each of the following requirements is met:

1. An ESF function is initiated when necessary.

2. The required protection system instrumentation coincidence logic is maintained.

3. Sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance.

Allowable values specified ensure that violation of the Safety Limits for the reactor core and RCS is prevented during normal operation and A00s, and the consequences of accidents are acceptable. Only the Allowable Values are specified for each ESFAS function. The allowable value is specified such that the analytical limit assumed in the safety analysis is conservative including all applicable setpoint uncertainties. (continued) Approved Design Meterini- 16A Tech Spec Bases rege B 3.3-107

Syst m 80+ oestan controlorcument ESFAS Instrumentation B 3.6.5 BASES SURVEILLANCE SR 3.3.5.5 (continued) i REQUIREMENTS identical for both the RPS and ESFAS, this is the same Surveillance performed for the RPS in SR 3.3.1.14. The CHANNEL FUNCTIONAL TEST for proper operation of the l bypass permissives is critical during plant heatups because ' the operating bypasses may be in place prior to entering. .. ; MODE 3, but must be removed at the appropriate points during plant startup to enable the ESFAS Function. Consequently, just prior to startup is the appropriate time to verify bypass function OPERABILITY. Once the operating bypasses are removed, the bypasses must not fail in such a way that the associated ESFAS Function is inappropriately bypassed. ' This feature is verified by SR 3.3.5.2. The allowance to conduct this surveillance within 92 days of startup is based on the reliability analysis presented in (Ref. 9). bypasses mu,0nce st not failthe in operating such a waybypasses that theare removed, the associated trip Function gets inadvertently bypassed. This feature is verified by the trip Function CHANNEL FUNCTIONAL TEST, SR 3.3.5.2. Therefore, further testing of the bypass removal function after startup is unnecessary. REFERENCES 1. M Section 7.3.

2. 10 CFR 50, Appendix A.

l

3. Reference Removed l

4. IEEE Standard 279-1971.

5. M Chapter 15.

6. 10 CFR 50.49.

7. (Setpoint Report).

8. @ Section 7.2.

9. [ Surveillance Interval Analysis]

10. Response Time Testing Acceptance Criteria. I 4*rmalD=6" Anotww TEA T=h spec sma re,e a 2.2-sn

I I Svatem 80+ l Deslan ControlDocument l ESFAS Logic and Manual Initiation )

B 3.3.6 l f

l BASES i : BACKGROUND logic either locally at the maintenance and test panels or dd ' (continued) remotely via the operator's module. The bypass status is ,, i available for display at the local maintenance and test to6ic G M^'#8 panels, remote operators modules, and DPS. ' M - ( e- % .,e d)- ACTUATION LOGIC wg The ESFAS Actuation Logic consists of a selective two-out-of-four logic for each ESFAS function. The inputs to the ACTUATION LOGIC are the LCL outputs from ; the appropriate local coincidence logics. The initiation circuits also contain a time delay (TD). The TD functions as a noise filter. It accomplishes this filter action by monitoring the continuous presence of an input for a minimum . period of time. If the signal is present for the required l time, the signal is transmitted to the initiation relay. - Test capability is also provided. i The initiation circuit is designed to fail-safe (i.e., in a ' trip condition). This will result in a partial trip (I of !

4) in the selective 2-out-of-4 ESFAS- actuation logic. The i partial trip will be alarmed the same as a full ESF trip and actuation and will be indicated by the DIAS and DPS; the partial trip cannot be bypassed. If the initiation circuit ,

fails in an undesired condition the failure will be promptly detected and alarmed via the automatic test function. Since the actuation functions in the ESF-CCS work in a selective coincidence logic, this is considered a degraded condition and a technical specification LCO will apply. CESSAR-DC Section 7.3 (Ref. I) describes ACTUATION LOGIC in detail. COMPONENT CONTROL LOGIC The COMPONENT CONTROL LOGIC is used to actuate the individual ESF components which are actuated to mitigate the consequences of the occurrence that caused the actuation. The ESFAS actuation and component control logics are physically located in four independent and geographically separate ESF-CCS cabinets. (continued) Anwowd Den 6" neareant.1sa rus seu suu p.,e a 3.312s

i e Svatem 80+ n-& controlDocument i i ESFAS Logic and Manual Initiation l B 3.3. l add ,, j BASES s A ! APPLICABLE 1. SafetyIn.iectionActuationSianal(SIAS)[Je># 4 - j~ SAFETY ANALYSES (continued) reactor coolant system and actuates components for

emergency cooling. SIAS also actuates containment p spray pumps. SIAS is also initiated by a loss of power to two or more measurement channels.

l -

2. Containment Sorav Actuation Sianal (CSAS) -

CSAS actuates containment spray, preventing j containment overpressurization during large break

;                                                              LOCAs, small break LOCAs, and MSLBs or feedwater line i                                                               breaks (FWLBs) inside containment. CSAS is initiated i                                                               by high containment pressure. CSAS is also initiated

by loss of power to two or more measurement channels.

j 3. Containment Isolation Actuation Sional (CIAS) . CIAS ensures acceptable mitigating actions during j large and small break LOCAs, and MSLBs inside containment or FWLBs either inside or outside containment. CIAS is initiated by low pressurizer pressure or high containment pressure. CIAS is also i initiated by loss of power to two or more measurement j channels. i j 4. Main Steam Isolation Sianal (MSIS) 4 l MSIS ensures acceptable consequences during an MSLB i or FWLB (between the steam generator and the main j feedwater check valve), either inside or outside

containment. MSIS isolates both steam generators if either gene *ator indicates a low pressure condition i

or if a hign Containment pressure condition exists. { This prevents an excessive rate of heat extraction

;                                                            and subsequent cooldown of the RCS during these                            -

q events. 5, 6. Emeraency Feedwater Actuation Sianal (EFAS) ' EFAS consists of two steam generator specific signals (EFAS-1 and EFAS-2). EFAS-1 initiates emergency feed 2 to SG #1 and EFAS-2 initiates emergency feed to SG

#2.

1 (continued) ) i ,

_ _ :Denkrr Meterial- 16A Tech Spec Bosee Pope 6 3.3-132

    . _ _ _ _ _ _ _ _ _ .                                                                     -                                                 -i

Sv' tem 80+ Deslan ControlDocument ESFAS Logic and Manual Initiation B 3.3.6 BASES SURVEILLANCE ACTUATION LOGIC Test (continued) REQUIREMENTS signals and the state of the actuation channel is automatically detected. 1 SR 3.3.6.2

A selective group test on each division of ACTUATION LOGIC and COMPONENT CONTROL LOGIC is performed to verify the OPERABILITY of each selective group. The [I8] month frequency is based on Reference 2 and the fact that some components cannot h tested at power since their actuation might lead to plant trip or equipment damage. ESFAS selective group testing is performed by an operator in the control room. This testing overlaps the PPS automatic testing of the ESF-CCS selective two-out-of-four coincidence logic and 1.ncludes complete testing of the ESFAS through to the actuation of the components. The components for each ESFAS are grouped. Testing is conducted one group at a time, thus preventing the complete undesired actuation of an ESF system during testing. Since this testing causes coaponents to actuate, an ESFAS signal from the PPS will not be hapeded and the ESF system will proceed to full a.ctuation. SP_ 3,3 J,3 A CHANNEL FUNCTIONAL TEST is performed on each ESFAS MANUAL INITIATION CHANNEL and on each DIVERSE MANUAL ESF ACTUATION CHANNEL to ensure the Actuation push buttons are capable of l enabling the ACTUATION LOGIC (for the MANUAL ESF ACTUATION l CHANNEL) or COMPONENT CONTROL LOGIC (for the DIVERSE MANUAL < ESF ACTUATION CHANNEL) when needed. The [I8] month frequency is based on Reference 2. 1 REFERENCES I. II55AR-isy:, Section 7.3. 4 I

2. (Surveillance Interval Analysis]

4, wow ous e , meeuw. ssa run spec sua roue a 3.2.sss

System 80+ Deskin ControlDocument DG - LOVS

                                                                                                       ,                         B 3.3.7 BASES SURVEILLANCE                                                    SR 3.3.7.3     (continued)

REQUIREMENTS for instrument drift between successive surveillances, to ensure the instrument CHANNEL remains operational. Measurement error determination, setpoint error determination, and calibration adjustment must be performed consistent with the plant specific setpoint analysis. The CHANNEL shall be left calibrated consistent __ with the assumptions of the current plant specific setpoint analysis. The as found and as left values must be recorded and l reviewed for consistency with the assumptions of the surveillance interval analysis. The requirements for this review are outlined in Reference (6). . l The setpoints, as well as the response to a Loss of Voltage l and Degraded Voltage test, shall include a single point

verification that the trip occurs within the required delay time as shown in Reference 1. The Frequency is based upon the assumption of an [18] month calibration interval for the determination of the magnitude of equipment drift in the setpoint analysis. i REFERENCES 1. T255AR% Chapter 8. 4 --

2. G 55AR- 4 Chapter 15. '

3. [Setpoint Report).

4. IEEE Standard 603-1980.

5. 10 CFR 50, Appendix A, GDC 21.

6. [ Surveillance Interval Analysis.]

l i W De*4Fr niesend- 16A rech see sesee pay,a 3,3.y53

4 System 80+ Deslan ControlDocument APS B 3.3.8 B 3.3 INSTRUMENTATION B 3.3.8 Alternate Protection System Instrumentation [ [ BASES r BACKGROUND The Alternate Protection System (APS) augments the Reactor Protective System to address 10 CFR 50.62 (Ref. 1) requirements for the reduction in risk of Anticipated Transients Without Scram (ATWS) and the use of ATWS .. Mitigating Systems Actuation Circuitry (AMSAC). The APS is designed to initiate a reactor trip for all J Anticipated Operational Occurrences (A00s) which cause an i overpressurization of the Reactor Coolant System with a I concurrent failure of the Reactor Protective System (RPS) to trip the reactor. These increasing pressure A00s include the following events ~ for System 80+:

Loss of Offsite Power
Loss of Load,
Loss of Condenser Vacuum,-
CEA Withdrawal From Low Power Conditions, i
Chemical and Volume Control System Malfunctions, and .
Loss of Normal Feedwater.

Of the above listed A00s, the Loss of Condenser Vacuum is the limiting event (i.e., the event with the lowest I Pressurizer Press - High trip setpoint and the highest Steam Generator Level - Low Emergency Feedwater Actuation System (EFAS) setpoint.] CESSAR-DC Sections 15.2, 15.4, and 15.5 (Ref. 4) provde a complete description of the above listed A00s. The APS design includes an Alternate Reactor Trip Signal (ARTS) and Alternate Feedwater Actuation Signal (AFAS) that are separate and diverse from the Plant Protection System (PPS). The ARTS equipment provides a simple, yet diverse mechanism to significantly decrease the possibility of an ATWS and the AFAS provides added assurance that an ATWS event could be mitigated if it were to occur. (continued) 0M~k w ou6n neeeuw tsA rech spec sua rene s 2.2su b

System 80+ oestan contrat occument APS B 3.3.8 l BASES l SR 3.3.8.3 'l SURVEILLANCE (continued) REQUIREMENTS

2. Electrical and mechanical instrument and equipment uncertainties must be considered.

However, harsh environment uncertainties need not be included for the AFAS setpoint analysis. 4 Mon i CHANNEL CALIBRATION shall find measurement errors are withir

the acceptance criteria specified in Reference 3 ~ b -fl-l The Frequency is based upon operating experience and R;9F consistency with the typical industry refueling cycle and is justified by the assumption of an [18] month calibration interval for the determination of the magnitude of equipment drift. REFERENCES 1. 10 CFR 50, Appendix A

2. CE3^,'Tv W Section 7.7 _z

3. [Setpoint Report) l

4. ( E 4 4= iC Sections 15.2, 15.4, and 15.5 m l

l l I 1 APoreved Des &n Materiel- 16A Tech Spec Bases Page B3 3.3-160 l

4 l Sv' tem 80 + Deslan Control Document i CRIFS B 3.3.9 i l BASES (continued) i l REFERENCES 1. -Ct. A Chapter 15. 4 *

2. (Setpoint Report]

i 3. 10 CFR 50, Appendix A, GDC 19. ! 4. [ Surveillance Interval Analysis.] i f: i i I i i t l 1 ) l i i s i 1 i

l i i 2 i Amoreved Desigrs Meteriel- 16A Tech Spec Bases page g 3.3.ggy 4

1 Syntem 80+ Deslan Control Document Containment Bypass Instrumentation SGTR B 3.3.10

  ' BASES a

SURVEILLANCE - SR 3.3.10.3 (continued) - j REQUIREMENTS ! i At this unit, CHANNEL CALIBRATION shall find measurement l _ errors are within the values specified in Reference 4. _ ! The Frequency is based upon operating experience and j consistency with the typical industry refueling cycle and is justified by the assumption of an [18] month calibration . interval for the determination of the magnitude of equipment drift. i g { REFERENCES 1. 10 CFR 50 Appendix A. 'l , i x M E5;A;- L Chapter 12. & j- 2. , 2 [3. CES3".", LSection 19.8, " Shutdown Risk Evaluation."] e

4. [Setpoint Report). -

! t i i l l i I i 4 i

    %.2 Dubn nichwW 16A Tuh Spec Buu                                             pape g 2,31y l

Syntem 80+ Design C?ntrol Document . PAMI ! B 3.3.11 BASES (continued) 4 , REFERENCES 1. -CEOSAR-;;. Chapter 7, " Instrumentation and Controls" =ci -

and 15 " Accident Analysis".: 2. Regulatory Guide 1.97.; 3. NUREG-0737, Supplement 1.

l i 4. System 80+ Procedure Guidelines _ 4 l 5. IEEE Standard 279,1971, " Criteria for Protection l Systems for Nuclear Power Generating Stations," April j 5, 1972. I I I i i i i i i ) i i4 I i i i 1 l ) i Approved Design MeterW- 16A Tech Spec Roses pay,g 3,3.gyo 4 4

Syntem 80+ Deslan ControlDocument i Remote Shutdown Instrumentation and Controls ! B 3.3.12 i i BASES I i a l SURVEILLANCE SR 3.3.12.2 (continued) j REQUIREMENTS 1 performed from'the control room, Vital Instrumentation i Equipment Rooms, and Remote Shutdown Room, as appropriate. { This will ensure that if the control room becomes

inaccessible, the plant can be brought to and maintained in i MODE 3 from the RSP. The [18] month Frequency is based on j the need to perform this Surveillance under the conditions ..
that apply during a plant outage and the potential for an i unplanned transient if the Surveillance were performed with

! the reactor at power. Operating experience demonstrates i that Remote Shutdown Instrumentation and Controls control f CHANNELS seldom fail to pass the Surveillance when performed at a Frequency of once every [18] months. l j SR 3.3.12.3 . ! CHANNEL CALIBRATION is a complete check of the instrument j CHANNEL including the sensor. The Surveillance verifies i that the CHANNEL responds to measured parameter with the j necessary range and accuracy. j The [18] month Frequency is based on the need to perform i this Surveillance under the conditions that apply during a

plant outage and the potential for an unplanned transient if l the Surveillance were performed with the reactor at power.

4 i SR 3.3.12.3 has been modified by a Note stating that Neutron l detectors are excluded from the CHANNEL CALIBRATION. i 1 l j REFERENCES 1. 10 CFR $0, Appendix A, GDC 19. i 2. CE556 -- Chapters 7 " Instrumentation and Controls," - - l and 15 " Accident Analysis." 4 e AnwendDen&n Atatede! 16A Tech Spec Page B 3.3-196 2

_ _ . . _ . _ _ __.._._._ _. . _ . ~ .._ _ . _ __ __ -_ - . .._- _ - _ . _ _ _ . _.

~ Syntem 80+ _% convolDocument Logarithmic Power Monitoring Channels B 3.3.13

) BASES SURVEILLANCE SR 3.3.13.3 - *

REQUIREMENTS j (continued) SR 3.3.13.3 is the performance of a CHANNEL CALIBRATION. A
CHANNEL CALIBRATION is performed every (18] months. The i Surveillance is a complete check and readjustment of the logarithmic power CHANNEL from the preamplifier input through to a remote display. The Surveillance verifies.that .

the CHANNEL responds to a measured REQUIREMENTS parameter __ .(

;                                        within the necessary range and accuracy.                   CHANNEL                                         l

CALIBRATION leaves the CHANNEL adjusted to account for i j instrument drifts between successive calibrations to ensure j that the CHANNEL remains operational. Measurement error determination, setpoint error determination, and i calibration adjustment must be performed consistent with the
plant specific setpoint analysis. The CHANNEL shall . be left
calibrated consistent with the assumptions of the current
plant specific setpoint analysis.

i j This SR is modified by a Note to indicate that it is not i necessary to test the detector, because generating a meaningful test signal is difficult; the detectors are of simple construction, and any failures in the detectors will be apparent as change in CHANNEL output. This test interval j is the same as that employed for the same CHANNELS in the other applicable MODES. 4 4 j REFERENCES 1. 10 CFR 50, Appendix A, GDC 13. 3 } 2. 6 Chapter 7 and Chapter 15. l : 1 1 Y i i i a 4 I I j? Deefpn Aneseniel 16A Tech Spu Run Pope B 3.3 201 i _ , _ .. , - - _ .~. _

System 80+ c::::., contrat Document :

Reactor Coolant Monitoring - Instrumentation-

B 3.3.14 l

! BASES i SURVEILLANCE SR 3.3.14.S (continued)

REQUIREMENTS '

reading accurately within specified tolerances prior to I j placing them in service and declaring them OPERABLE for ! J inventory monitoring. l 1

This calibration requirement of [60 days) is intended to be

!                                       more restrictive than the normal calibration frequencies                                    __

specified for a particular MEASUREMENT CHANNEL that may -

l 'already be in service for other plant monitoring purposes. i f REFERENCES 1. jE5 g L Appendix 19.8A, Shutdown Risk Evaluation c 1 __r_. _.

i. 2. NUREG-1449, Shutdown and Low-Power Operation at Commercial Nuclear Power Plants in the United States, j Draft Report, February 1992.

4 i l l i l j i I l 1 1 i i i 1 1 W W N W

I U I W W S**** & R 3,3 313

System 80+ Deslan controlDocument RCS Pressure, Temperature and Flow limits .' B 3.4.1 BASES i , SURVEILLANCE SR 3.4.1.4 (continued) - - REQUIREMENTS The SR is modified by a Note which states the SR is only required to be met (24) hours after reaching 90% RTP. The Note is necessary to allow measurement of the flow rate at normal operating conditions at power in MODE 1. The Surveillance cannot be performed in MODE 2 or below, and will not yield accurate results if performed below 90% RTP. , REFERENCES 1. M Chapter 15. 4 2 Approved Deelgrr Metodel.16A Tech Spec Boses page a 3,s g

   ,. -         - .- ..._    ..      - . . - ~ .-          .     ,   --      .            -      . - . - . .        -. - _ __

System 80+ Deslan controlDocument i i RCS Minimum Temperature for Criticality l j B 3.4.2 j l BASES i SURVEILLANCE SR 3.4.2.1 (continued) REQUIREMENTS delay criticality so the LCO will not be violated. The 30 minute time is frequent enough to prevent inadvertent , violation of the LCO. l While this Surveillance is required whenever the reactor is critical and temperature is at or below [550*F], in practice - the Surveillance is most appropriate during the period when the reactor is brought critical. Because the operator would likely verify average RCS temperature more often than required by this Surveillance, it is less restrictive than normal operating practice. REFERENCES 1. M Chapter 15. % i d i t Anwoved Doelps hiehwiel 16A Tech Spec Buu pay, y up

          . . .           -                       -.     -     -   _       -      . - - - . .                -   =.

    - _ . . - - - . . - - . . - - .              - . - . _ _ _ . - - - ~ .           - --       _ . - . . --                -   .- .          .

1 l 1 Syntem 80+ Deslan controlDocumast l l RCS P/T Limits j B 3.4.3

BASES j LC0 of the vessel to become more pronounced), and the -

l (continued) consequences also depend on the existences, sizes and l orientations of flaws in the vessel material. Although i vessel failure is not an expected outcome of a violation,

the possibility for failure exists. p i ,

j ! i APPLICABILITY The RCS P/T limits provides a definition of acceptable ; operation for prevention of non-ductile failure that is in > l accordance with 10 CFR 50 Appendix G (Ref.1). Although the- ) P/T limits were developed to provide guidance for operation

during heatup and cooldown (MODES 3, 4, and 5) or ISLH 1 testing, their Applicability is at all times in keeping with j the concern for non-ductile failure. At all times is j defined to be any condition with fuel in the reactor vessel.

The limits do not apply to the pressurizer. ! However, during MODES I and 2, other Technical i Specifications provide limits for operation that can be more

restrictive than or can supplement the P/T limits. These j other LCOs include LCO 3.4.2, "RCS Minimum Temperature for

, Criticality," and LCO 3.4.1, "RCS Pressure, Temperature, and 4 Flow Limits." SL 2.1, safety limits for pressure and temperature and maximum pressure, also provides operational restrictions. In MODE 6, with the reactor vessel head detensioned or removed, the capability for violating the P/T curves does not exist, however the potential for violating the temperature rate-of-change limit remains. Furthermore, in MODES 1 and 2, operation is above the j temperature range of concern for non-ductile failure. As l

such, stress analyses have been developed in accordance with i normal maneuvering profiles such as power ascension.

? l The actions of this LCO consider the premise that a j violation of the limits occurred during normal plant i l maneuvering. Severe violations caused by abnormal } transients, which may be accompanied by equipment failures, i may also require additional actions based on emergency j operating procedures. 1 i j j Obf f

!                                                                                                              (continued)    hM:

Approved Dubn nieteniel.16A Tech Spec Buu pope g 3,417

i. - . . . . . _ .

Sv t m 80 +

_ Des 19n Control Document RCS Loops - MODES I and 2 i B 3.4.4 4 i BASES l ! i - jAPPLICABILITY The decay heat production rate is much lower than the full - - (continued) power heat rate. As such, the forced circulation flow and ; heat sink requirements are reduced for lower, noncritical { MODES as indicated by the LCOs for MODES 3, 4, 5, and 6. 4 l Operation in other MODES is covered by: J LCO 3.4.5, "RCS Loops-MODE 3"; ] LCO 3.4.6, "RCS Loops-MODE 4"; - a LC0 3.4.7, "RCS Loops-MODE 5 (Loops Filled)"; 1 LC0 3.4.8, "RCS Loops-MODE 5 (l. oops Not Filled)"; : I LCO 3.9.4, " Shutdown Cooling System (SCS) and Coolant j Circulation-High Water Level" (MODE 6); and i LC0 3.9.5, " Shutdown Cooling System (SCS) and Coolant i Circulation-Low Water Level" (MODE 6). ACTIONS Ad If the required number of loops are not in operation, the Required Action is to reduce power and bring the plant to i MODE 3. The action lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits. It should be noted that the reactor will trip and place the plant in MODE 3 as soon as the Reactor Protection System senses less than four RCPs operating. The six hours allowed is a reasonable time based on , operating experience to reach MODE 3 from full power without challenging safety systems. SURVEILLANCE SR 3.4.4.1 REQUIREMENTS l This SR requires verification of the required number of loops in operation and reactor coolant circulation every 12 l i hours to ensure that forced flow is providing heat removal. The 12 hour interval has been shown by operating practice to be sufficient to regularly assess degradation and verify . operation within safety analysis assumptions. The i verification may be performed by checking RCPs in operation %9e i and RCS flew and temperature indications, t d ,2 Ig V i 9 5 I ta continued) e u 2 g w A ' n .-10 w . n ,a r ,s,.. e , nu

                                                                                                  %)qY.y        M

A

'   System 80+                                                                                I
                                             ..Deslan controlDocument ll                                         RCS Loops - MODES 1 and 2 B 3.
! BASES (continued)

REFERENCES 1. M Chapter 15. < ' 4 Y j 1 J t

i k i i I i r .i ) i k i i I 1 i e i h I' i 1 1 't t f. i i h .l 1 i I 1 d MswendDuy MeterW 16A Tech Sm Bues Page B 3.4 26 i

1 Syntem 80+ oestan controlDocument . RCS Loops - MODE'3 B 3.4.5 lB3.4 REACTOR COOLANT SYSTEM (RCS) i iB 3.4.5 RCS Loops - MODE 3 , 1 1 'l ' l BASES BACKGROUND The primary function of the reactor coolant in MODE 3 is ; ! removal of decay heat and transfer of this heat, via the

steam generators, to the secondary plant fluid. The i secondary function of the reactor coolant is to act as a -

carrier for soluble neutron poison, boric acid. . l In MODE 3, reactor coolant pumps (RCPs) are used to provide i forced circulation heat removal during heatup and cooldown.

The MODE 3 decay heat removal requirements are low enough that a single RCS loop with one RCP running is sufficient to remove core decay heat. However, [two] RCS loops are

! required to be OPERABLE to satisfy single failure criteria. j Only one RCP need be OPERABLE to declare the associated RCS - j loop OPERABLE. i ! Reactor coolant natural circulation is not normally used, ! but is sufficient for core cooling. However, natural

circulation does not provide turbulent flow conditions.

i Therefore, boron reduction in natural circulation is j - prohibited because mixing to obtain a homogeneous concentration in all portions of the RCS cannot be ensured. .i APPLICABLE Analyses have shown that the rod withdrawal event from MODE jSAFETYANALYSES 3 with one RCS loop in operation is bounded by the rod r withdrawal initiated from MODE 2. t l Failure to provide heat removal may result -in challenges to ! a fission product barrier. The RCS loops are part of the

primary success path which functions or actuates to prevent j or mitigate a design basis accident or transient that either .

j assumes the failure of, or presents a challenge to, the 1, integrity of a fission product barrier.. RCS loops - MODE 3 satisfy Criterion 3 of the NRC Policy Statement. li I e women uneaw 1sa recs spec saa (continued) rouea u n k i

1 ! I l ' Sv' tem 80+ i Desian controlorcument ! RCS Loops - MODE 4 i 1 B 3.4.6 i } BASES j LCO In MODES 3, 4, and 5, it is sometimes necessary to stop all - * ! (continued) RCP or SCS pump forced circulation (i.e., change operation l from one SCS division to the other, perform surveillance or i startup testing, perform the transition to and from SCS, or to avoid operation below the RCP minimum NPSH limit). The , time period is acceptable because natural circulation is j adequate for heat removal or the reactor coolant temperature . can be maintained subcooled, and boron stratification affecting reactivity control is not expected.

i An OPERABLE RCS loop consists of at least one OPERABLE RCP

and a steam generator that is OPERABLE in accordance with

! the Steam Generator Tube Surveillance Program and has the j minimum water level specified in SR 3.4.6.2. ! Similarly, for the SCS, an OPERABLE SCS division is composed of the OPERABLE SCS pump (s) capable of providing forced flow to the SCS heat exchanger (s). RCPs and SCS pumps are - I OPERABLE if they are capable of being powered and are able { to provide flow if required. APPLICABILITY In MODE 4, this LCO applies because it is possible to remove core decay heat with either the RCS loops and steam generators or the SCS. Operation in other MODES is covered by:

LC0 3.4.4, "RCS Loops - MODES 1 and 2";

LCO 3.4.5, "RCS Loops - MODE 3"; LCO 3.4.7, "RCS Loops - MODE 5 (Loops Filled)";

LCO 3.4.8, "RCS Loops - MODE 5 (Loops Not Filled)";

;                              LCO 3.9.4,        " Shutdown Cooling System (SCS) and Coolant
!                                                Circulation - High Water Level" (MODE 6); and
}                              LC0 3.9.5,        " Shutdown Cooling System (SCS) and Coolant j

1 Circulation - Low Water Level" (MODE 6). i l l 61ivh Lidd,

 ;                                                                                                            % ey e s i

Appf0M MW A6000d0f

1M rOCh $90C S0000 90 g3 i_ _ ._ _

System 80+ Denkm ControlDocument RCS Loops - MODE 5 (Loops'Not Filled) B 3.4.8 i j BASES (continued) ! . .i SURVEILLANCE SR 3 . 4. 8.1 - - 'l REQUIREMENTS 1- This SR requires verification of the required SCS division in operation every 12 hours to ensure forced flow is providing heat removal. Verification of SCS operation is

performed by flow rate, temperature, or pump status i monitoring. The 12 hour Frequency has been shown by 1 i operating practice to be sufficient to regularly assess )

degradation and verify operation within safety analyses' l

assumptions.

h 1 SR 3.4.8.2 l i .l j Verification that the required number of SCS divisions are I

OPERABLE ensures that redundant paths for heat removal are

! available and that additional SCS divisions can be placed in j operation,~ if needed, to maintain decay heat removal and - , reactor coolant circulation. Verification is performed by j verifying proper breaker alignment and indicated power available to the required SCS pumps. The Frequency of seven i days is considered reasonable in view of other i administrative controls available and has been shown to be j acceptable by operating experience. 1 i

SR 3.4.8.3

! Verification of the correct breaker alignment and indicated

power available to the operable CS pump ensures that the l redundant CS pump will be able to remove heat from the RCS j in the event of a power failure to the operating SCS-division. The Frequency of [24] hours is based on operath j experience.

l REFERENCES 1. ^;00"t, ; Chapter 5. y

2. C ;;JJ T' , Chapter 19.

l i l 1 l l . i i } AnwendDulgn AtewW= 16A Tnh Seu Bene Page B 3. w

System 80+ Deslan CmntrolDocument LTOP System B 3.4.11 BASES SURVEILLANCE SR 3.4.11.2

                                                                                                                                                                   -l REQUIREMENTS (continued)                                       The pressurizer manway must be verified open for relief protection. The Frequency of every 12 hours is sufficient to verify compliance within the safety analysis assumptions.

The Surveillance is modified by a Note only requiring performance when manway is being used for LTOP. SR 3.4.11.3 Surveillance Requirement 3.4.11.3 is the performance of a setpoint calibration every (18 months). The setpoint calibration for the LTOP ensures that the SCS relief valves will be actuated at the appropriate RCS pressure by verifying the accuracy of the valve lift pressure. The Frequency of (18 months] is based on a typical refueling - cycle and industry-accepted practice. i l REFERENCES 1. 10 CFR 50, Appendix G, " Fracture Toughness i l Requirements."

2. Generic Letter 88-11, "NRC Position on Radiation Embrittlement of Reactor Vessel Materials and its Impact on Plant Operation."

3. 6 Chapter 15. ;

4. Generic Letter 90-06.

l ww u ww ssa rus sm sua r.o.a u s1

i

                                                                                                                                                    ^& ControlDocument                            i l Sv~ tem 80+

RCS Operational LEAKAGE I

B 3.4.12 i B 3.4 REACTOR COOLANT SYSTEM (RCS) lB3.4.12 RCS Operational LEAKAGE 4 BASES 4 . . - _ . . . . -. _ - - _ - _ . _ _ . . - _ - . . . - . _ _ - _ - - . . ~ - - . . . - . ~ - - . . _ - - _. - -.. --- ...---- -. I ! BACKGROUND Components that contain or transport the coolant to or from a the reactor core make up the RCS. Component joints are made l by welding, bolting, rolling, or pressure loading, and ] valves isolate connecting systems from the RCS. _ l j During plant life, the joint and valve interfaces can j produce varying amounts of reactor coolant LEAKAGE, through

either normal operational wear or mechanical deterioration.

! The purpose of the RCS Operational LEAKAGE LCO is to limit [ system operation in the presence of LEAKAGE from these

sources to amounts that do not compromise safety. This LCO

{ specifies the types and. amounts of LEAKAGE. 1 j 10 CFR 50, Appendix A, GDC 30 (Ref.1), requires means for - detecting and, to the extent practical, identifying the j source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems. i The safety significance of RCS LEAKAGE varies widely I depending on its source, rate, and duration. Therefore, j detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary. Quickly separating the' i identified LEAKAGE from the unidentified LEAKAGE is !l necessary to provide quantitative information to the ! operators, allowing them to take corrective action should a , j leak occur detrimental to the safety of the facility and the- { public. I A limited amount of leakage inside containment is expected i from auxiliary systems that cannot be made 100% leaktight. l Leakage from these systems.should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS LEAKAGE detection. This LCO deals with protection of the reactor coolant j pressure boundary (RCPB) from degradation and the core from

inadequate cooling, in addition to preventing the accident j analysis radiation release assumptions from being exceeded.

The consequences of violating this LC0 include the possibility of a loss of coolant accident (LOCA). l i j

4provedDubn hineerW 16A Tuh Spec Bene (continued)

Pope E 3.4-62

@v[

I 1 _ _ _ _ _ _ _ _ _ _ _ _ ___._ . . _ . _. ~ ,_ _

System 80+ Deslan convolDocumart RCS Operational LEAKAGE B 3.4.12 j BASES e SURVEILLANCE SR 3.4.12.1 (continued) - ' i REQUIREMENTS The 72 hour Frequency permits a reasonable interval for trending of LEAKAGE while recognizing the relative importance of early leak detection in the prevention of 1 accidents. A Note under the Frequency column states that this SR is required to be performed during steady state operation. Steady state operation is required to perform a - proper inventory balance; calculations during maneuvering are not useful and the Surveillance is not required unless steady state is established. For purposes of LEAKAGE determination by inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank level, constant makeup and letdown and reactor coolant pump seal injection and return i flows. Pressure Boundary LEAKAGE would be detected more quickly by the LEAKAGE detection systems referenced in LCO 3.4.14, "RCS LEAXAGE Detection Instrumentation". - SR 3.4.12.2 This SR provides the means necessary to determine SG ' OPERABILITY in an operational MODE. The requirement to demonstrate SG tube integrity in accordance with the Steam Generator Tube Surveillance Program emphasizes the importance of SG tube integrity, even though this Surveillance cannot be performed at normal operating conditions. REFERENCES 1. 10 CFR 50, Appendix A, GDC 30. i

2. Regulatory Guide 1.45, May 1973. .

l 3. - CEssnn-sit, Chapter 15. _eg

4 1

?,^,^ ^& Dee> A000ersel 16A Tech Spec Bases Pop B 1667 l

5 _. _ _ . 7 - . , - -

Shtem 80+ Deslan control Document RCS LEAXAGE Detection Instrumentation B 3.4.14 BASES t , SURVEILLANCE SR 3.4.14.3. SR 3.4.14.4 and SR 3.4.14.5 - REQUIREMENTS (continued) These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the l instrument channel, including the instruments located inside : containment. The Frequency of [18] months is a typical refueling cycle and considers channel reliability. ~ Operating experience has shown this Frequency is acceptable. REFERENCES 1. 10 CFR 50, Appendix A, Section IV, GDC 30.

2. Regulatory Guide 1.45, U. S. Nuclear Regulatory Commission.

3. % Chapter 5. 4___

l l n hyweved Design Meteniel 16A Tech Spec Beser pay, a 3.4.go

Sv? tem 80L Deslan ControlDocument RCS Specific Activity B 3.4.15 BASES SURVEILLANCE SR 3.4.15.1 (continued) REQUIREMENTS This Surveillance is applicable in MODES 1, 2, and in MODE 3 with RCS average temperature at least 500*F. The Frequency of 7 days considers the unlikelihood of a gross fuel failure during the time. SR 3.4.15.2 ~ This Surveillance is performed to ensure iodine levels remain within limits during normal operation and following fast power changes when fuel failure is more apt to occur. The 14 day Frequency is adequate to trend changes in the - activity level considering that gross activity is monitored every 7 days. The Frequency between two and six hours following a power change ;t 15% RTP within a one hour period is established because iodine spikes during this time following fuel failure. Samples at other times would provide inaccurate results. SR 3.4.15.3 A radiochemical analysis for E determination is required to be performed every 184 days (six months) with the plant. operating in MODE 1 with equilibrium conditions. These requirements for E determination directly relate to the LCO and are required to verify plant operation within the specified gross activity LCO limit. The radiochemical , analysis for E is a measurement of the average energies per I disintegration of isotopes with half lives > 15 minutes, I excluding iodines. The Frequency of 184 days is based on the fact that E does not change rapidly during operation. The Frequency of 184 days recognizes E does not change rapidly. This SR has been modified by a Note that indicates sampling is required to be performed within 31 days after 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last suberiticci for at least 48 hours. This ensures the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event. QM N% l

h. ()g

- o . 7 ,, (cg,"!ii

Sy' tem 80+ Design ControlDocument i RCS Specific Activity l B 3.4.15 l I 1 BASES (continued) l l

                                                                                                 - .l 3EFERENCES                 1.      10 CFR 100, " Determination of Exclusion Area, Low               '

Population Zone, and Population Center Distance," USNRC, 1973.

2. -ibwe-se, Chapter 15.

4 I l l l l l l l peroved Desip, Meteniel.16A Tech Spec Bases Page B 3.4-86

Wstem 80 + Design Control Document Reactor Coolant Gas Vent System B 3.4.17 ASES JRVEILLANCE SR 3.4.17.3 - EQUIREMENTS l l(centinued) Verifying that the pressure instrument root valves are open ensures that line pressure between can be monitored. The 18 month Frequency is based on accessibility during the refueling cycle. ! I l SR 3.4.17.4 " Verification of correct breaker alignment and valve position indications ensures that valves can be operated when required and valve position can be monitored. The Frequencv of seven days has been shown to be acceptable by operatinc experience. (FERENCES 1. 7E5:nmaA Chapter 6. :- I l i l foved Design Material- 16A Toch Spec Bases Page B 3.4-93

f Syntem 80+ Deslan ControlDocument 1 i l l Rapid Depressurization Function ' 8 3.4.18 BASES (continued)

SURVEILLANCE SR 3.4.18.1 - -

1 REQUIREMENTS Verifying that the pressure instrument root valves are open j ensures that line pressure between the globe and gate valves can be monitored. The [18 month] Frequency is based on 4 accessibility during the refueling cycle. SR 3.4.18.2 " r l Cycling each vent valve through at least one complete cycle i verifies the RDF valves will function when necessary. The i Frequency of [18 rionths) is based on a typical refueling

!                               cycle and is an industry accepted practice.

l SR 3.4.18.3 4 Periodic verification of the correct valve position 2 indication in the control room for all RDF valves ensures that the valves are properly aligned and that the position indicators are functioning properly. A Frequency of [12] hours is accepted by industry practice, and has been shown ) to be acceptable by operating experience. i SR 3.4.18.4 Verification of correct breaker alignment and power availability to the valve indicators ensures that valves can be operated when required, and valve position can be i' monitored. The Frequency of seven days is accepted industry practice, and has been shown to be acceptable by operatin-g experience. l REFERENCES 1. ^eoom- 4 Chapter 6. 4 - i 1 l I i l Approveef Des / prs Material 16A Tech Spec Bases Page R 3.496

Svotem 80+ oesian controlDocument

. Vent Paths - REDUCED RCS INVENTORY Operations B 3.4.19 5 k g, 3,9 BASES (continued) 6,,, , s

1. 0;; Q Appendix 19.8A, Shutdown Risk Evaluation 4

i iREFERENCES 1 r- j Dele E4 "" s Repe rS i 'I .. J l t } l. 3 1 I i l i I, 1 i i

  ' Appesved Design Material 16A Tech Spec Bases                                       Page B 3.499 i

4 i Sy tem 80+ oeslan controlDocument i 1 j SITS B 3.5.1 1 4 j BASES (continued) 1 REFERENCES 1. IEEE Std. 279-1971, Criteria For Protection Systems

for Nuclear Power Generating Stations.

!' 2. 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants.

3. Q Chapter 6. 4

4. 'tR$ARJtEChapter 15. 1 l

) 5. NUREG-1366, " Improvements to Technical Specification ! Requirements." I i i 1

1 } i 1 i i i l i i i 4 i i i I - J l i I 4 i i i i f l Anoroved Design Materiel 16A Tech Spec Bases page g 3.5.g0

1 System 80+ Deskn ControlDocument SIS - Operating B 3.5.2 1 i BASES SR 3.5.2.5 St%VEILLANCE i REQUIREMENTS (continued) [ Discharge head at design flow is a normal test of SI pump performance required by Section XI of the ASME Code. The Frequency for such tests is a Code requirement. Such inservice inspections detect component degradation and incipient failures.] SR 3.5.2.6 and SR 3.5.2.7 j These SRs demonstrate each automatic SIS valve actuates to ! its required position on an actual or simulated Safety ! Injection Actuation Signal (SIAS) and that each SIS Pump

starts on receipt of an actual or simulated SIAS. The(18]

month Frequency is based on the need to perform these Surveillances under the conditions that apply during a plant , outage and the potential for unplanned transients if the Surveillances were performed with the reactor at power. The [18] month Frequency is also acceptable based on consideration of the design reliability (and confirming operating experience) of the equipment. The actuation logic j is tested as part of the Engineered Safety Feature Actustion System (ESFAS) testing, and equipment performance is

monitored as part of the Inservice Testing Program.

SR 3.5.2.8 i Periodic inspections of the IRWST Holdup Volume Tank ensures that it is unrestricted and it stays in proper operating condition. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during an outage, on the need to have access to the location, and on the potentfal for unplanned transieni.s if the Surveillance were performed with the reactor at power. This Frequency is sufficient to detect abnormal degradation and is confirmed by operating experience. 1 REFERENCES 1. 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Plants.

2. $Afl% Chapter 6.

(continued) AnwovedDesign Meteriel 16A Tech Spec Reser Pope B 3.51g

Syntem 80+ outon conma Document SIS - Shutdown B 3.5.3 BASES (continued) ACTIONS U * ' ) With only one SI Pump UPERABLE, the unit is not prepared to respond to a LOCA. The I hour Completion Time to restore at

least two SIS trains to OPERABLE status ensures prompt action is taken to restore the required cooling capacity.

1 B.I.1. B.1.2. and B.2 __ The plant must be placed in a condition in which the LCO does not apply if SIS cannot be returned to OPERABLE status within the associated completion time. An RCS level of 2 (120'0"] (the top of the reactor vessel flange) will provide a minimum water inventory in the event of a LOCA. In addition, the reduction of RCS temperature to <135'F will provide a reduction in clad temperature. The 24 hour Completion Time limits the time the plant is subject to conditions where the LCO is applicable.

SVRVEILLANCE SR 3.5.3.1 REQUIREMENTS-The applicable Surveillance descriptions from Bases LCO 3.5.2 apply. REFERENCES 1. 10 CFR 50.46, Acceptance Criteria for Emergency Core j Cooling Systems for Light Water Nuclear Power Plants.

2. M Chapter 6.

3. 10 CFR 50, Appendix A, GDC 35 - Emergency Core Cooling System. \

4. NRC Memorandum R. L. Bayer to V. Stello, Jr.,-

                                   " Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

5. NRC Information Notice No. 87-01, RHR Valve Misalignment Causes Degradation of ECCS in PWRs, January 6, 1987.

6. M Section 19.8A, " Shutdown Risk Evaluation"

         ~               . ,n r- e     e                                                 ew,
  - =                                  . - .          -

I 4 i . Sv-tem 80+ oestan control Documart l 1 IRWST l B 3.5.4 3 1 l BASES (continued) j -. , REFERENCES 1. 10 CFR, Appendix A, GDC 35 - Emergency Core Cooling l Systems.

2. @ Chapter 6.

3. % Chapter 15.

i 1 i l4 a _ i j 3 l i .l d l , i i b - l i l j J l l 1 i i Aporewed Deeipts Materiel- 16A Tech Spec Bases Page B 3.5-28

l SCtem 80 + Design ControlDocument , 14 ASES 6.5

URVEILLANCE SR 3.5.5.1 (continued): EQUIREMENTS The periodic verification is required every [18] months, since access to the TSP baskets is only feasible during outages, and normal fuel cycles are scheduled for 18 months.

Operating experience has shown this Surveillance Frequency acceptable due to the margin in the volume of TSP placed in the containment. . __ SR 3.5.5 2 Testing must be performed to ensure the solubility and buffering ability of the TSP after exposure to the containment environment. A representative sample of [33.8] grams of TSP from one of the baskets in containment is submerged in 1.0 i 0.05 gallons of water at a boron concentration of [4400] ppm and at the standard temperature of 25 i 5'C. Without agitation, the solution pH should be raised to 2 7.0 within 4 hours. The representative sample weight is based on the minimum required TSP weight of [23,939] kilograms, which at manufactured density corresponds to a minimum volume of [926] cubic feet and maximum possible post LOCA recirculated water volume of [708,316] gallons, normalized to buffer a 1.0 gallon sample. The boron concentration of the test water is representative of the maximum possible boron concentration corresponding to the maximum possible post LOCA HVT recirculated water volume. Agitation of the test solution is prohibited, since an adequate standard for the agitation intensity cannot be specified. The test time of 4 hours is necessary to allow time for the dissolved TSP to naturally diffuse through the sample solution. In the post LOCA HVT, rapid mixing would occur, significantly decreasing the actual amount of time before the required pH is achieved. This would ensure compliance with the Standard Review Plan requirement of a pH 2 7.0 by the onset of recirculation after a LOCA. m (FERENCES None. l l ' proved Design Material 16A Tech Spec Bases pay, g 3,$.33

Syntem 80+ Deslan ControlDocument CFS B 3.5.6 BASES (continued) . SURVEILLANCE SR 3.5.6.1 REQUIREMENTS Verification of proper valve position ensures that the correct CFS alignment is maintained. Misalignment of these valves could affect the OPERABILITY of the IRWST. The 31 ' day Frequency is based on operating experience. SR 3.5.6.2 The CFS valves are stroke tested in accordance with the Inservice Inspection and Testing Program, and Section XI of the ASME Code. A test interval based on the refueling outage Frequency was specified considering it is prudent that this Surveillance be performed only during a plant outage. Testing of the HVT flooding valves requires that the manual valves located upstream be closed to prevent the - flow of water from the IRWST to the HVT. Closing the manual valve is not practical during plant power operations because - containment entry is required. Based on the low probability of a severe accident requiring use of the CFS and the redundancy provided by the design, a refueling outage interval for stroke testing the HVT flooding valves is considered acceptable. The reactor cavity flooding valve are to be stroke test at the same at the same interval as in the HVT flooding valves. i REFERENCES 1. SECY-90-016 of 1/12/90,

Subject:

Evolutionary Light i Water Reactor (LWR) Certification Issues and their ;

Relationship to Current Regulatory Requirements. l

, 2. . Atha-i rPChapter 3. , l 3. M ':[,, Chapter 6. ! -t _ ! 4. - -CE55% Chapter 19. w = ouir,uno-w. m rm sm sua r.,, , s, e.sy . l i

        . _ _ . _ . _ _ , -      -.       .- . . _ -                ,   ~ _ _ _ . .

   - ~ . . - - .                    _ - - -          . . - - . - _ _ . . - .         - - . .                           . . . -     - - - _            - _ .

l System 80+ Desinn ControlDocument i Containment i B 3.6.1

         .B 3.6 CONTAINMENT SYSTEMS j          B 3.6.1 Containment                                                                                                                   -

BASES I BACKGROUND The containment vessel, including all its' penetrations, is a low leakage steel shell which is designed to withstand the postulated Loss of Coolant Accident (LOCA) or a Main Steam Line Break (MSLB) while limiting the postulated release of -

radioactive material to within the requirements of 10 CFR 100 (Ref. 1). Additionally, the containment and shield l building provide shielding from the fission products which i may be present in the containment atmosphere following i accident conditions.

i The containment vessel is a 200-ft. diameter spherical steel shell with a wall thickness of approximately one and three-

quarter inches. This containment-shell is supported by a. .

2 spherical depression in an intermediate floor of the shield

building. The containment.is enclosed by a reinforced l concrete cylindrical shield building with 'a hemispherical

! dome. An annular space exists between the steel containment vessel and the shield building. The internal structure is a group of reinforced concrete structures that enclose the reactor vessel and primary system. The internal structure provides biological ) shielding for the containment interior. The internal structure concrete base rests inside the lower portion of ! the containment vessel sphere. . The primary shield wall encloses the reactor vessel and provides protection for the vessel from internal missiles. The primary shield wall provides biological shielding and is designed to withstand the temperatures and pressures i following a LOCA. In addition, the primary shield wall i provides structural support for the reactor vessel. The - i primary shield wahl is a minimum of six feet thick. The secondary shield wall (crane wall) provides supports for the polar crane and protects the steel containment vessel from internal missiles. In addition to providing biological y g shielding for the coolant loop and equipment, the crane dMp wall also rovides structural upport for pipe supports /r straints and platfor s t arious d evels. -The i l (continued)

       .^;; .JDen&n Metenlof 16A Tech Spec Basee                                                          page g g,g.g

I System 80+ D ~ sign Control Document - G Contai t B 3.6.1 fASES j [ - BACKGROUND crane Owall Sis a right cylinder with an inside diameter of (continued) 130 feet and a height of 118 feet from its base. The crane wall is a minimum of four feet thick. Containment piping penetration assemblies provide for the passage of process, service, sampling and instrumentation j pipe lines into the containment vessel while maintaining containment OPERABILITY. The shield building provides - - biological shielding and controlled release of the annulus atmosphere under accident conditions, and environmental missile protection for the containment vessel and Nuclear Steam Supply System. The inner steel containment and its penetrations establish the leakage limiting boundary of the containment. Maintaining the containment OPERABLE limits the leakage of fission product radioactivity from the containment to the - environment. Loss of containment OPERABILITY could cause site boundary doses, in the event of a DBA, to exceed values given in the licensing basis. SR 3.6.1.1 leakage rate requirements comply with 10 CFR 50, Appendix J (Ref. 4), as modified by approved exemptions. The isclation devices for the penetrations in the containment boundary are a part of the containment leak tight barrier. To maintain this leak tight barrier:

a. All penetrations required to be closed during accident conditions are either:

1. capable of being closed by an OPERABLE automatic containment isolation system, or

2. closed by manual valves, blind flanges, or de-activated automatic valves secured in their closed positions, except as provided in LC0 3.6.3, " Containment Isolation Valves".

b. Each air lock is OPERABLE except as provided in LC0 3.6.2, " Containment Air Locks".

i l l l l l (continued) ppmved Design Matada!- 16A Tech Spec Bases Page B 3.6-2

System 80+ Desinn control Document Containment B 3.6.1 BASES (continued) APPLICABLE The safety design basis for the containment bis.that the

                                                                                                                         /                        '

SAFETY ANALYSES containment must withstand the loadings of the limiting DBA without exceeding the design leakage rate. l The DBAs which result in a release of radioactive material )' within containment are a Loss of Coolant Accident (LOCA), a Main Feedwater Line Break (MFLB), and a Control Element Assembly (CFA) ejection accident (Ref. 2). In the analysis -- of each of these accidents, it is assumed that the containment and containment shield building are OPERABLE at event initiation such that the majority of the release of , fission products to the environment is controlled by the I rate of containment leakage. In addition, for the above I accidents, it is assumed that the containment low volume ) purge is operating at event initiation. Isolation of the purge will be automatic or manual depending upon the pressure transient associated with the analyzed accident. . The containment was designed with an allowable leakage rate of [0.5) percent of the containment volume per day (Ref. 3). ' This leakage rate, used in the evaluation of offsite doses resulting from accidents, is defined in 10 CFR 50, Appendix J (Ref. 4) as La: the maximum allowable containment leakage rate at the calculated maximum peak containment pressure j (Pa) following a DBA. The calculated maximum peak l containment pressure (48.1) psig was obtained from a-[0%) I power MSLB DBA. The containment internal design pressure is (53.0) psig. The allowable leakage rate represented by La i forms the basis for the acceptance criteria imposed on all containment leak rate testing. Satisfactory leak test results are a requirement for the establishment of containment OPERABILITY. The acceptance criteria applied to accidental releases of radioactive material to the environment are given in terms of total radiation dose received by a member of the general public who remains at the exclusion area boundary for two hours following onset of the postulated fission product ' release. The limits established in Reference 1 are a whole body dose of 25 Rem or a 300 Rem dose to the thyroid from iodine exposure, or both. (continued) Arenmdone, ueraw. Ien rea spee sua re,e a 2.e.s

I j

! 3 System 80+ o ston controlDocument 1

Containment B 3.6.1 BASES (continued) ] . .

REFERENCES 1. 10 CFR 100.11, " Determination of Exclusion Area, Los Population Zone, and Population Center Distance."

2. N Chapter 15. 4 i 3. -4fSS31R4C Chapter 6. 2 v

J 4. 10 CFR 50, Appendix J, " Primary Reactor Containment - l Leakage Testing for Water-Cooled Power Reactors." i 1 4 e I e i ( ) l 4 l a 1 ? 1 4 3 i' l i i 1 wmr outen staw- 1sA rech spec suu rare s 2.s-s

i System 80+ Deslan contror Document Containment Air Locks i B 3.6.2 i r l BASES jSURVEILLANCE SR 3.6.2.1 (continued)

~ REQUIREMENTS l                      of a DBA. Note 2 has been added to this SR requiring.the i                      results to be evaluated against the acceptance criteria of.

! SR 3.6.1.1. This ensures that air lock leakage is properly ! accounted for in determining the overall containment leakage

rate.

SR 3.6.2.2 I The air lock door interlock is designed to prevent

simultaneous opening of both doors in a single air lock.

Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post-accident i containment. pressure [48.1 psig), closure of either door

will support containment OPERABILITY. Thus, the door -

interlock feature supports containment OPERABILITY while the , air lock is being used for personnel transit into and out of containment. Periodic testing of this interlock l demonstrates that the interlock will function as designed and that simultaneous opening of inner and outer door will not inadvertently occur. Due to the purely mechanical nature of this interlock, and given that the interlock i mechanism is only challenged when containment is entered, l this test is only required to be performed upon entering j containment but is not required more frequently than every j 184 days. 4 l / REFERENCES 1. 10 CFR 100.11, " Determination of Exclusion Area, l i Population Zone and Population Center Distance." l

2. --ff5$Ai>9fer Chapter 15. <

{ 3. CC';';&% Chapter 6. 4 _ ! 4. 10 CFR 50, Appendix J, " Primary Reactor Containa j Leakage Testing for Water-Cooled Power Reactors.' t t f y;;2 Denke neesaw. IsA Tech spea suee rare s 2.s.14 f

\A

(.-- - - . -.. - . - . _ .. - .. - - _ . - - . . _ - -- 1 l System 80+ Deslan ControlDocument i Containment Isolation Valves l B 3.6.3 l B 3.6 CONTAINHENT SYSTEMS i ! B 3.6.3 Containment Isolation Valves . . j . . j BASES 4 BACKGROUND The containment structure serves to contain radioactive l material which may be released from the reactor core i following a Design Basis Accident (DBA), such that offsite i radiation exposures are maintained within the requirements - j of 10 CFR 100 (Ref. 1). The containment isolation valves

form part of the containment pressure boundary and provide a l means for fluid penetrations not serving accident

, consequence limiting systems to be provided with two

isolation barriers that are closed on a Containment Isolation Actuation Signal (CIAS). These isolation devices are either passive or active (automatic). Manual valves, de-activated automatic valves secured in their closed I position (including check valves with flow through the valve-

{ secured), blind flanges, and closed systems are considered { passive devices. Check valves, or other automatic valves

!                           designed to close without operator action following an i

accident, are considered active devices. Two barriers in series are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the accident analysis. One of these barriers may j be a closed system. ! Automatic containment isolation occurs upon receipt of a l high containment pressure signal or Safety Injection i Actuation Signal (SIAS). The CIAS closes automatic

containment isolation valves in fluid penetrations not 3 required for operation of engineered safeguards systems in l order to prevent leakage of radioactive material. Other penetrations are isolated by the use of valves in the closed 4

position or blind flanges. As a result, the containment { isolation valves (and blind flanges) help ensure that the ~ j containment atmosphere will be isolated in the event of a i release of radioactive material to containment atmosphere

from the RCS following a DBA. OPERABILITY f the Containment Isolation Valves (and blind fla es) ensures containment integrity is maintained during a cident

conditions.

1 k )

i (continued) j - AnnrevedDenho AteterW.16A Tech See Bases page a 2.s.1g

i System 80+ Deslan ControlDocument Containment Isolation Valves B 3.6.3 BASES (continued) t 1 REFERENCES 1. 10 CFR 100.11, " Determination of Exclusion Area, Low j Population Zone, and Population Center Distance."

2. ,7.5;Aa,%, Chapter 6. <

, 3. 4:00AR-Bir, Chapter 15. q l 4. Generic Issue (GI B-20), " Containment Leakage Due to

Seal Deterioration."

! 5. NRC Generic Item B-24, " Purge Valve Reliability." l 6. 10 CFR 50, Appendix J, " Primary Reactor Containment i Leakage Testing for Water-Cooled Power Reactors." ] i l l a ! i i i \ j J a i Wuwd Design Meredat.16A Tech Spec Basee

                                                                                       !*9083.630

System 80+ oestan control Document i Containment Pressure

B 3.6.4 BASES (continued)

,1 REFERENCES 1. 10 CFR 100,11, " Determination of Exclusion Area, Lov Population Zone, and Population Center Distance." )

2. --GE55A D4 Chapter 6.

3. AE55AR-iKp, Chapter 15.

j; , i e 4 1 J j i i 1 1 A J } I i I a ] I 1 4 i 1 l 4 4 1 4 5 i i d i t i l 4preved Design Morenia!- 16A Tech Spec Banen page g 2.g.34 4

System 80+ Desian ControlDocument Containment Air Temperature B 3.6.5 BASES (continued) REFERENCES 1. 10 CFR 100.11, " Determination of Exclusion Area, Li Population Zone, and Population Center Distance."

2. 'E55Aji:ss. Chapter 6.

3. CE^^^^ "" Chapter 15.

l I I i i AnwoMDesign Meterial.16A Toca Spec Bases pay, a 3 g.3g 1

System 80+ , _ o ston controlDocument Containment Spray System B 3.6.6 BASES (continued) REFERENCES 1. Ci3; 4 Chapter 6. 4

2. CE;ta ^z - Chapter 15. =:Q

3. 10 CFR 100.11. " Determination of Exclusion Area, Loi Population Zone, and Population Center Distance."

4. -CS88Mh5r Appendix 19.8A, Shutdown Risk Evaluation I I l

Anwevenf Design Matend = 16A Tech Spec Basee p,y, y 3,g.45 l

i System 80+ Design Control Document

.                                                                      Hydrogen Analyzers B 3.6.7 4

BASES i 5URVEILLANCE SR 3.6.7.1 (continued)

REQUIREMENTS l

The 92 day Frequency has been shown to be acceptable through operating experience and is consistent with the

;                        recomendations of NUREG-1366 (Ref. 3).

SR 3.6.7.2 - i Performance of a CHANNEL CALIBRATION on the Hydrogen i Analyzers using sample gases ensures the OPERABILITY of the

!                        analyzers is maintained. A typical CHANNEL CALIBRATION includes a minimum of two data points to verify accuracy of

the analyzers over the range of interest. The sample gases

;                        used for performing the Surveillances are nominally % by i                         volume hydrogen 2 [0.98] and s [1.02] (balance nitrogen),

and nominally % by volume hydrogen 2 [3.92] and s [4.08] - (balance nitrogen). The lower hydrogen flamability limit is assumed as 4.0% by volume hydrogen in air or steam-air j atmospheres. Therefore,. calibration with these sample gases ! helps ensure accurate information regarding containment hydrogen concentrations up to and including the flamability limit is available to operators following a LOCA. The 18 i month Frequency has been shown to be acceptable through operating experience and is consistent with the recomendations of NUREG-1366 (Ref. 3). I l

REFERENCES 1. Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in containment Following a Loss-of-Coolant Accident, Revision 2, November 1978. I j 2. -@t*8Mt4C, Chapter 6. "

3. NUREG-1366, " Improvements to Technical Specification j Surveillance Requirements." ,

4 i 4 i

   &vedDesign Metodel 16A Tech Spec Bases                                         Page B 3.6-50

System 80+ o ston contrat Document Shield Building B 3.6.8 BASES

                                                                                                                                                                   . s SURVEILLANCE                                                                         SR  3. 6.M (continued)

REQUIREMENTS The (18] month Frequency to verify the required negative pressure in the shield building is consistent with Regulatory Guide 1.52 (Ref 4) guidance for functional testing of the ability of the AVS to " pull down" the required negative pressure every (18] months. REFERENCES 1. TE55A33iles, Chapter 3. 1 -

2. 6 Chapter 6. <

3. 10 CFR 100.11, " Determination of Exclusion Area, Lov Population Zone and Population Center Distance."

4. Regulatory Guide 1.52 (Rev. 02), " Design, Testing ano Maintenance Criteria for Post Accident Engineered Safety Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants."

l ? Anwed Design Metada! 16A Tech Spec Baser pay,g 3,g.54

i_ Syst m 80+ Desian controlDocument I Annulus Ventilation System B 3.6.9 i l i BASES i ACTIONS 8.1 and B.2 * ' ] i (continued) If the inoperable AVS cannot be restored to OPERABLE statt d/os e v )v i within the required Completion Time, the plant must be m placed in a MODE in which the LCO does not apply. To .. i achieve this status, the plant must be brought to at least deM_*" j 2 MODE 3 within 6 hours and to MODE 5 within 36 hours. % e allowed Completion Times are reasonable based on operating - experience to reach the required plant conditions from full 4 power conditions in an orderly manner and without ) challenging plant systems. 1 ) SURVEILLANCE SR 3.6.9.1 ] REQUIREMENTS

Operating each AVS division, with the heaters operating, -

ensures that all divisions are OPERABLE and that all

associated controls are functioning properly. It also ensures that blockage, fan or spotor failure, or excessive

! vibration can be detected for corrective action. (Operation 1' with the heaters on for 2 10 continuous hours eliminates moisture in the filtration units. Experience from filter j testing at operating units indicates that the 10 hour period

is adequate for moisture elimination in the filtration 1 units.] The 31 day Frequency was developed considering the i known reliability of fan motors and controls, the two i

4 division redundancy available, and the iodine removal capability of the Containment Spray System. j SR 3.6.9.2 i 4 This SR verifies that the required AYS filter testing is performed in accordance with the [ Ventilation Filter Testing l Program (VFTP)]. Filter tests are in accordance with ) Regulatory Guide 1.52 (Ref. 3 . j j of HEPA filter performance, i m)nimum system flow rate, andThe VFTP inc i the physical properties of the activated carbon (general use i and following specific operations). Specific test i frequencies and additional information are discussed in detail in the VFTP. i 4 1 (continued) !! l Anonowenf Design Meterial 16A Tech Spec Basee Page 3 3.6 50 i

l l system 80+ oesion contrat oocument Annulus Ventilation System B 3.6.9 l BASES SURVEILLANCE .SE_, 'a . 6. 9. 3 I REQUIREMENTS I (continued) The automatic startup ensures that each AVS division responds properly. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at-power. Operating experience has shown that -- these components usually pass the Surveillance when performed at the [18] month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint. Furthermore, the SR interval was developed considering that the AVS equipment OPERABILITY is demonstrated at a 31 day Frequency by SR 3.6.9.1. SR 3.6.9.4 . The AVS division flow rate is verified s [18,000) cfm to ensure that the flow rate through the filters -is not excessive. The [18] month on a STAGGERED TEST BASIS Frequency is consistent with the Regulatory Guide 1.52 (Ref. 3) guida REFERENCES 1. - CE55A; u Chapter 6.

2. iE55im-jar, Chapter 15. 4 I

3. Regulatory Guide 1.52 (Rev. 02), " Design, Testing ;

Maintenance Criteria for Post Accident Engineered Safety Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled Nuclear Power Plants." l l l i l M:. 2 Dukn wtaw.16A Tech spec suu y, y 1s.go

                   ..                       .-~      .             .   .

  . . _ . _         . . . _ _ . ~ . . . .        . _ . _     . _ _ _ _    . _ .       . . . _ - . _ . _ .  .-.        ._   . . . . _ _ -          .

l l M 80+ Desian ControlMwt l Containment Penetrations - REDUCED RCS INVENTORY Operations 8 3.6.11 l j BASES ACTIONS R.d (continued) 4 If Action A.1 has not been completed within the [6] hours, then the RCS level must be restored to > [EL-117' 0"] within j (6) hours of Action A.1 not being met. i l \ \

i

! SURVEILLANCE SR 3.6.11.1 i j REQUIREMENTS 1 i This SR verifies that each required containment building i j penetration is in its required status every [12 hours). ' l This ensures that fission products will'not escape j cortair. ment in a quantity greater than assumed in the safety i analysis. 1 i 1 SR 3.6.11.2 . i i This SR verifies each containment purge and exhau.st valve i actuates to its isolated position on an actual or simulated i 1 actuation signal. The (18 month] Frequency maintains consistency with similar ESFAS testing requirements and h j been shown to be acceptable through operating experience. Deld q "dESSN#' C { REFERENCES 1. @ Appendix 19.8A, Shutdown Risk Evaluation ~ g 04 , j . R _J i < r I ! l 1 ) 1 i j w % me.w.rs4r=*s,-su. ,,,,,3., j

i System 80+ Deslan Control Document i - MSSVs B 3.7.1 ! l

B 3.7 PLANT SYSTEMS j B 3.7.1 Main Steam Safety Valves (MSSVs) _ ,

BASES i

 ,  BACKGROUND                         The Main Steam Safety Valves (MSSVs) mainly provide over-i                                       pressure protection for the secondary system. In doing so, l                                       the MSSVs also provide protection against overpressurizing the reactor coolant pressure boundary by providing a heat                                             _

sink for removal of energy from the Reactor Coolant System i (RCS) if the preferred heat sink, provided by the condenser j and Circulating Water system, is not a/ailable. Five Main Steam Safety Valves, (ten per steam generator) are j located on each Main Steam Line, outside Containment,

upstream of the Main Steam Isolation Valves, as described in i CESSAR-DC Chapter 5 (Ref.1). The MSSVs' rated capacity

passes the full steam flow at 102% RATED THERMAL POWER i

' (RTP)(100 + 2% for inttrument error) with the valves full c>M, open. This meets the requirements of the ASME Code (Ref. 2) as described in the Over-pressure Protection Report, N

WAppendix 5. A. The MSSV design includes staggered G. ,D L j setpoints, as shown in Table 3.7.1-2, so that only the

! number of valves needed will actuate. The maximum system 1 overpressure is calculated based on maximum allowable tolerance on the MSSV setpoint. Staggered setpoints reduce . the potential for valve chattering because of insufficient 1 steam pressure to fully open all valves following a turbine- ! reactor trip. ! G R lk ! APPLICABLE

                           /(he'Ref. design

2) andbasis for the MSSVs limits secondary comes tofrom system pressure <; 110% the ofASME Code j SAFETY ANALYSES design pressure when passing 100% of design steam flow.

This design basis is more than sufficient to cope with any .' Anticipated Operating Occurrence (A00) or accident. - i considered in the Design Basis Accident and Transient . . Analysis. For most analyzed events, RCS pressure remains

below the setpoint of the Pressurizer Safety Valves (PSVs),

j or, at most, cause only a short opening of the PSVs. i The events that challenge the MSSVs' relieving capacity, and j thus RCS pressura, are those characterized as Decreased Heat Removal svents, and are presented in Section 15.2 J % i (continued) i i Annrowd Dee&n nietodel 16A Tech Spec Seeen Pope B 3.71 , l

System 80+ Deslan Control Document MSSVs B 3.7.1 BASES D.A[d n APPLICABLE -tt3322:ar (Ref. 4). Of these, the full power Loss of d.- D C *i SAFETY ANALYSES Condenser Vacuum (LOCV) event is the limiting A00. A LOCV (continued) isolates the turbine and condensor, and terminates normal feedwater flow to the steam generators. Before delivery of Emergency Feedwater (EFW) to the steam generators, RCS pressure reaches s [2,726] psia. This peak pressure is less than 110% of the design pressure of 2,500 psia, but high enough to actuate the PSVs. The maximum secondary pressure ~ during the LOCV event is [1273] psia, which is less than 110% of secondary design pressure of 1200 psia. The limiting accident for peak RCS pressure is the full l power feedwater line break, inside Containment, with a loss of offsite power. Water from the affected steam generator is assumed to be lost through the break with minimal additional heat transfer from the RCS. With heat removal limited to the unaffected steam generator, - the reduced heat transfer causes an increase in RCS temperature and the resulting RCS fluid expansion causes an increase in pressure. The RCS pressure increases to s [2,798] psia, with the PSVs providing relief capacity. The maximum secondary pressure during the feedwater line break event is s [1273] psia, which is less than the rated i capacity of 110% of the design pressure of 1200 psia. The MSSVs satisfy Criterion 3 of the NRC Policy Statement. LC0 The LCO requires all MSSVs to be OPERABLE in compliance with the ASME Code.0peration with less ti.an the full number of MSSVs requires limitations on allowable THERMAL POWER (to meet ASME Code requirements) and adjustment to the Reactor Protective System Trip Setpoints. These limitations are addressed in Table 3.7.1-1 and Required Actions A.1 and A.2. An MSSV is considered inoperable if it fails to open upon demand. The OPERABILITY of the MSSVs is defined as the ability to open within the setpoint tolerances, relieve steam generator over-pressure and to re-seat when pressure has been reduced. The OPERABILITY of the MSSVs is determined by periodic surveillance testing in accordance with the Inservice Testing Program. { { (continued) prowd oute, neesarw. sea run spee snee rene a 2.7 2

Syntem 80+ Deslan controlDocument MSSVs B 3.7.1 3ASES LCTIONS B.1 and B.2 - - (continued) The plant must be placed in a MODE in which the requirement does not apply if the MSSVs cannot be restored to OPERABLE status in the associated Completion Time. This is done by placing the plant in at least MODE 3 in six hours and in MODE 4 in [12 hours). The allowed Completion Times are reasonable based on operating experience to reach the ~ ~ required MODES from full power operation without challenging plant systems, iURVEILLANCE SR 3.7.1.1 tEQUIREMENTS This SR demonstrates the OPERABILITY of the MSSVs. The ASME Code Section XI (Ref. 5) requires that safety and relief valve tests be performed as required by ASME/ ANSI OM-1-1987 - - (Ref. 3). According to Reference 3, the following tests are required for MSSVs:

a. Visual examination,

b. Seat tightness determination,

c. Set pressure determination (lift setting),

d. Compliance with owner's seat tightness criteria, and

e. Verification of the balancing device integrity on .

balanced valves. The ANSI /ASME standard requires testing all valves every five years, with a minimum of 20% of the valves tested every 24 months. Surveillance requirements are specified in the Inservice Testing Program which encompasses Section XI of ,I the ASME Code. ASME Code provides.the activities and frequencies necessary to satisfy the requirements. SR 3.7.1.1 is modified by a Note that allows entry into and - operation in MODE 3 prior to performing the SR. This is to allow testing of MSSVs at hot conditions. The MSSVs may be either bench tested, or tested in-situ at hot conditions using an assist device to simulate lift pressure. If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure. '3 A i ,Q 4 (continued) m 5~y y 5 % q perewer cuer neurw- 1sA ruh seu seu S v ( rope s 2.7.s l4 1

i i Syntem 80+ Desian control Document

MSSVs B 3.7.1 l

j BASES (continued) REFERENCES 1. t3331tsta8G Chapter 5. *

2. ASME Boiler and Pressure Vessel Code, Section III, Article NC-7000, " Overpressure Protection" Class 2 1

,                                   Components.                                                            I 1

1

3. ANSI /ASME OM-1-1987, " Requirements for Inservice l Performance Testing of Nuclear Power Plant Pressure Relief Devices." '

4. N Chapter 15. 4

5. ASME Boiler and Pressure Vessel Code, Section XI,

Article IWV-3500, " Inservice Tests - Category C i Valves."

i i i i i 3 4 1,

AAprowd Deskrr Material 16A Tech Spec Bases p,y y 3, y g

l l - System 80+ Deelan ControlDocument

i MSIVs j B 3.7.2 B 3.7 PLANT SYSTEMS llB3.7.2 l Main Steam Isolation Valves (MSIVs) , l ! BASES j BACKGROUND The Main Steam Isolation Valves (MSIVs) isolate steam flow I from the secondary side of the steam generators following a j high energy line break. MSIV closure terminates flow from j the unaffected (intact) steam generator. _ i

!                                                                       One MSIV is located in each Main Steam line outside, but

] close to, containment. The MSIVs are downstream from the

!                                                                       MSSVs, ADVs and Emergency Feedwater Pump turbine's steam j                                                                        supplies to prevent their being isolated from the steam j                                                                        generators by MSIV closure. The MSIVs have bypass valves 2

that allow the warming of the downstream main steam piping. 1 Closing the MSIVs isolates each steam generator from the ! other, and isolates the turbine, steam bypass system, and 3 other auxiliary steam supplies from the steam generators. - The MSIVs and MSIV Sypass Valves close on a Main Steam Isolation Signal (MSIS) generated by either low steam generator pressure, high containment pressure, or high steam j generator water level. The MSIVs and bypass valves fail close on loss of control or activation power. The MSIS also j actuates the Main Feedwater Isolation Valves to close. The j MSIVs and bypass valves may also be closed manually. - l A description of the MSIVs and bypass valves is found in Chapter 10 N (Ref.1). C be Y l The design basis of the MSIVs is established by the q , D Q,

    -APPLICABLE                                                       containment analysis for the large steam line break inside SAFETY ANALYSES                                                  containment (Ref. 2). It is also influenced by the accident

! analysis of the steam line break events presented in Chapter 15 ef4$e4lf!!5RR=D (Ref. 3). The design precludes the C f blowdown of more than one steam generator, assuming a single active component failure, i.e., the failure of one MSIV to , j close on demand. 1 !' The limiting case for the containment analysis is the hot 4 zero power steam line break inside containment with a loss of offsite power following turbine trip and failure of one l t i j (continued) i w omw, ne,ww. saa rus soee sua p ,, a 2.7 7 4

             . . . _ . ~ . . _ , - - .                                                          ,

, \ l 1 \ Syntem 80+ Deskn controlDocument l

MSIVs i B 3.7.2 I 1 BASES i SURVEILLANCE SR 3.7.2.1 (continued) -

                                                                                                ': (

! REQUIREMENTS i stroke exercise increases the risk of a valve closure with the unit generating power. As the MSIVs are not tested at power, they are exempt from the ASME Code, Section XI (Ref. j 4), requirements during operation in MODES I and 2. The Frequency for this SR is in accordance with the ~ j [ Inservice Testing Program or (18] months]. This (18] month Frequency demonstrates the valve closure time at least once ) per refueling cycle. Operating experience has shown that i these comoonents usually pass the SR when performed at the j (18] montb Frequency. Therefore, the Frequency is j J acceptable from a reliability standpoint. 3 This test should be conducted in MODE 3, with the unit at ! operating temperature and pressure, as discussed in the Reference 4 exercising requirements. This SR is modified by - i a Note that allows entry into and operation in MODE 3 and 4 { prior to performing the SR. This allows a delay of testing

until MODE 3, in order to establish conditions consistent t

with those under which the acceptance criterion was j generated. I 1 REFERENCES 1. MMMWW99 Chapter 10. " 4

2. e63322: 0E Chapter 6.; 3. lie 85mR40E-Chapter 15. %

l 4. American Society for Mechanical Engineers. Boiler an. 1 Pressure Vessel Code, Section XI, Inservice Inspection. Article IWV-3400 " Inservice Tests - Category A and B Valves." i j 5. 10 CFR 100, Reactor Site Criteria.

I i i i i i i } Anrend De& A0eterW 16A Tech Sne Been pope g 3.712 1

System 80+ Design ControlDocument MFIVs B 3.7.3 BASES BACKGROUND The MFIVs and a check valve inside containment is available I ' (continued) to isolate the feedwater line penetrating containment, and to ensure the consequences of events do not exceed the capacity of the containment heat removal systems. A description of the MFIVs is found in Chapter 10 sMNE

                      -CS$32fket (Ref.1).

x - APPLICABLE The design basis of the MFIVs is established by the analysis dDC SAFETY ANALYSES for the large steam line break (SLB). It is also influenced by the accident analysis for the large feedwater line break (FWLB). Closure of the MFIVs may also be relied on to terminate an excess feed event upon the generation of a MSIS on high steam generator level. Failure of a MFIV to close following an SLB, FWLB, or excess feedwater flow event can result in additional mass and energy to the steam generators contributing to cooldown. This failure also results in additional mass and energy releases following an SLB or FWLB event. The MFIVs satisfy Criterion 3 of the NRC Policy Statement. .C0 This LC0 ensures that the MFIVs will isolate feedwater flow ' to the steam generators. Following a feedwater or main stream line break, these valves will also isolate the nonsafety related portions from the safety related portions of the system. This LCO requires that two MFIVs in each feedwater line be OPERABLE. The MFIVs are considered OPERABLE when their isolation times are within limits, and are closed on an isolation actuation signal. Failure to meet the LC0 requirements can result in additional mass and energy being released to containment { following a steam or feed line break inside containment. If MSIS on high steam generator level is relied on to terminate an excess feedwater flow event, failure to meet the LC0 may , i result in the introduction of water into the main steam lines. l l wwend Design Meterial 16A Tech Spec Beses (continued) Page B 3.714

System 80+ oestan controloccument MFIVs B 3.7.3 BASES SURVEILLANCE SR 3.7.3.1 (continued) REQUIREMENTS ASME Section XI (Ref. 2) requirements during operation in MODES I and 2. l This SR ensures that the MFIVs are fully tested at least once per refueling cycle. The Frequency is in accordance with the (Inservice Testing Program or 18 months.] The - Frequency of (18] months is based on the refueling cycle and has been shown to be acceptable through operating experience. REFERENCES 1. ZEssa>06 Chapter 10.

2. American Society for Mechanical Engineers, Boiler and Pressure Vessel Code, Section XI, Inservice

! Inspection, Article IWV-3400 " Inservice Tests - Category A and B Valves." l l l l l Annrevent Dee# Motorte! 16A Tech Spec Besee Page 3 3.717

  .       .       , ~ - _ - - .             ~-                       .                    . . . .          -- _-_

Systrm 80+ Deskn CrntrolDocument EFW B 3.7.4 , BASES a BACKGROUND generator. The cavitating venturi restricts the magnitude -

                                                                                                                     'i (continued)                of the two pump flow as well as the magnitude of individual                           :

pump runout flow to the steam generator. A cross-connection is provided between each EFWST so that either tank can supply either division of EFW. Pump discharge crossover piping is provided to enhance system versatility during long-term emergency modes, such that a - single pump can feed both steam generators. Two normally locked closed, local manually operated isolation valves are provided for subdivision separation. One-hundred percent capacity is sufficient to remove decay a heat and cool the plant to shutdown cooling entry conditions at the design cooldown rate, [100*F/hr]. Fifty percent capacity is sufficient to remove decay heat but is insufficient to maintain the design cooldown rate. The diverse motive power of the two divisions meets the diversity requirement of BTP ASB 10-1 (Ref. 6). The EFW System is one of the systems required to meet GDC 34 and GDC 44 regarding the capability to remove decay heat and transfer it to an ultimate heat sink, in this case the

              -                 atmosphere.

An OPERABLE EFW System is required if the steam generators are to be considered OPERABLE. The EFW System is discussed in Chapter 10 sAmree99MWelt F , (Ref. 1). g,pC APPLICABLE The EFW System mitigates the consequences of any event with SAFETY ANALYSES a loss of normal feedwater. The EFW System provides at least the minimum required flow to the steam generators to - meet the design basis heat removal requirements. Following the event, the EFW System maintains adequate feedwater inventory in the steam generator (s) for heat removal and is capable of maintaining hot standby and facilitating a plant cooldown from hot standby to shutdown cooling system , initiation. (continued) i Anvuod Deskn nieteriel 16A Tech Spec Bosee page g 17.gg

System 80+ oesign controt Document EFW B 3.7.* BASES (continued) REFERENCES 1. ww ro Chapter 10. M -

2. NRC Generic Letter 88-03, " Resolution of Generic Safety Issue 93, " Steam Binding of Auxiliary feedwate i Pumps", February 17, 1988.

3. American Society for Mechanical Engineers, Boiler ano Pressure Vessel Code, Section XI, Inservice --

Inspection, Article IWV-3400 " Inservice Tests - Category A and B Valves."

4. 10 CFR 50, Appendix A, GDC 34 - Residual Heat Removal.

5. NUREG-0800, " Standard Review Plan", (SRP) Section 10.4.9, Rev. 2, " Auxiliary Feedwater System (PWR)."

6. Branch Technical Position ASB 10-1, " Design Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply Diversity for Pressurized Water Reactor Plants."

l Approved Design Afsterial- 16A Tech Spec Bases p,,, y 3. 7 24

Sv" tem 80+

Deslan controlDocument EFWST B ?.7.5 8 3.7 PLANT SYSTEMS iB3.7.5 Emergency Feedwater Storage Tank (EFWST) t . .

BASES iBACKGROUND The Emergency Feedwater Storage Tanks (EFWSTs) provide a

safety grade source of water for removing decay and sensible i heat from the Reactor Coolant System (RCS) during emergency phases of the plant. The EFWST provides a passive flow of ~

water by gravity to the emergency feedwater (EFW) pumps.

The EFW pumps supply this water tc the steam generators to t

remove heat from the RCS. The steam produced is released to the atmosphere by the main steam safety valves (MSSVs) or the atmospheric dump valves (ADVs). When the main steam isolation valves (MSIVs)'are open, the preferred means of heat removal is to discharge steam to the

condenser by the non-safety grade path of the turbine bypass valves. This has the advantage of conserving condensate 4

                                                                                                          -             ~

while minimizing releases to the environs. ! There are two EFWSTs, each tank provides suction head for one motor-driven and one steam-driven EFW pump. i A normal locked closed, local manually operated isolation valve is provided for each EFWST to provide separation. A line connected to a non-safety source of condensate is also provided with local manual isolation so that it can be manually aligned for gravity feed to either of the EFWSTs, should the EFWSTs reach low level before Shutdown Cooling g System entry conditions are reached. Each tank contains 100% of the total required water. supply. N ) A description of the EFWST is found in G888MutiLChapter 10 4 ! (Ref. 1). Because the EFWST is a principal component in removing residual heat from the RCS, it is designed to withstand earthquakes and other natural phenomena, as well b' as missiles which might be generated by natural phenomena. The water volume of each EFWST 2 [350,000] gallons is determined by the quantity required to achieve safe cold shutdown considering: (continued) armed neewe aneeww- rea reas spec seene reo, a s.1-2s

Syotem 80+ Deslan controlDocument EFWST B 3.7.5

ASES
ACKGROUND a. A main feedline break without isolation of EFW flow to - '

(continued) the affected steam generator for 30 minutes.

b. Refill of the intact steam generator.

c. Eight hours of operation at hot standby conditions.

d. Subsequent cooldown of RCS within six hours to conditions which permit operation of the Shutdown -

Cooling System.

e. Continuous operation of one reactor coolant pump.

At the end of this cooldown, the EFWST level must be sufficient to ensure adequate NPSH for the operating EFW pumps. The EFWST is one of the systems required to meet GDC 34 and GDC 44 regarding the capability to remove decay heat and transfer it to an Ultimate Heat Sink. PPLICABLE AFETY ANALYSES The EFWST provides cooling water to remove decay heat and cooldown the plant following all events in the accident P analysis, 'G143*Ca> Chapters 6 and 15. For anticipated 4

operating occurrences and accidents which do not affect the (L DC operability of the steam generators, the analysis assumption is generally 30 minutes at MODE 3, steaming through the MSSVs, followed by a cooldown to Shutdown Cooling entry conditions at the design cooldown rate. The N- Chapters 6 and 15 accident analysis does not form the basis for the EFWST volume as the events analyzed require less condensate than the design basis. The limiting event for the condensate volume is the large feedwater line break with a loss of offsite power. Single failures that also affect this event include; 1) the failure of the diesel generator powering the motor driven EFW pump to the unaffected steam generator (requiring additional steam to drive the remaining EFW pump's turbine); and 2) the failure of the steam driven EFW pump. These are not usually the limiting failures in terms of consequences for these events. (continued) oroved Design A4sterial- 16A Tech Spec Bases Page B 3.7-26

i System 80+ i Deslan ControlDocument i EFWST

B 3.7.5 4

l BASES 1 l APPLICABLE The EFWST satisfies Criterion 3 of the NRC Policy Statement *l 2 SAFETY ANALYSES because it is in the primary success path for all events in O which the steam generators are available for heat removal 7 4 (continued) l j fromtheRCS.p yo ] l s-l LC0 To satisfy accident analysis assumptions, the EFWST must

contain sufficient cooling water to remove decay heat for 30: minutes following a reactor trip from 102% RATED THERMAL j POWER and then cooldown the RCS to Shutdown Cooling entry i conditions, assuming a loss of offsite power and the most
' adverse single failure. In doing this it must retain sufficient water to ensure adequate NPSH for the EFW pumps during the cooldown, as well as to account for any losses i from the steam driven EFW pump's turbine, or before isolating EFW to a broken line. ,

j The specified usable volume of 2 [350,000] gallons is based i on holding the plant in MODE 3 for eight hours followed by a j cooldown to Shutdown Cooling entry conditions at l (100*F/ hour]. This bases is established by BTP RSB 5-1 (Ref. 2) and exceeds the volume required by the accident

analysis.

! APPLICABILITY The required condensate volume must be available whenever the steam generators provide the heat sink for the RCS. Once a cooldown commences, the condensate volume may be i reduced by using it for the cooldown. Proceeding with the

cooldown ensures that the plant can reach Shutdown Cooling l entry conditions on the available condensate inventory.

!                            In MODES 5 and 6 the steam generators are not required for I                            cooldown, and the inventory in the EFWST is not required.

i I I ! ACTIONS A.1 and A.2 4

If the EFWST level is not within the limit, the OPERABILITY of the other EFWST must be verified by administrative means j

within 4 hours.

i i l (continued) i Approwd Deelyn Aceteriel 16A Tech Spec Buu Page B 3.7 27

System 80+ '

                                                                                     &&n ControlDocument I

EFWST l ! .B 3.7.5 .il BASES

I i ACTIONS A.1 and A.2 (continued) - ' OPERABILITY of the other EFWST must include verification of ! the OPERABILITY of flow paths from the tank to the EFW i pumps, and availability of the required volume of water. i i

The EFWST level must be returned to OPERABLE status within 7 I days. The 4 hour Completion Time is reasonable, based on operating experience, to verify the OPERABILITY of the other EFWST. The 7 days Completion Time is reasonable, based on

i an OPERABLE EFWST being available. I { B.1 and B.2 I When a Required Action cannot be completed within the

Completion Time, a controlled shutdown should be commenced.

j Six hours is a reasonable time, based on operating l experience, to reach MODE 3 from full power conditions j without challenging plant systems. I j Continuing the plant shutdown begun in Required Action B.1, 4 [18] hours is a reasonable time, based on operating ! experience, to reach MODE 4 from full power conditions l without challenging plant systems. d i SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Checking the EFWST level 2 [350,000] gallons verifies that the EFWST contains the required volume of cooling water. Checking once per 12 hours is adequate because the operator > will be aware of plant evolutions which can affect the EFWST inventory between checks. { REFERENCES 1. - Chapter 10. 4 i 2. Branch Technical Position RSB 5-1, " Design Requirements for the Residual Heat Removal System." f l I 3 i 4 l i l a

;                          .-7                                                                     ,.,. , m, 1

1

Sv' tem 80+ Desian controlDocument Secondary Specific Activity l B 3.7.6 BASES SURVEILLANCE SR 3.7.6.1 (continued) - IEQUIREMENTS releases. It also serves to identify and trend any unusual isotopic concentrations which might indicate changes in reactor coolant activity or LEAKAGE. The [31] day Frequer.cy l allows the level of DOSE EQUIVALENT I-131 to be monitored, ' increasing trends to be detected, and appropriate action to be taken to maintain levels below the LC0 limit. REFERENCES 1. 10 CFR 100, " Site Dose Criteria".

2. .IEEERR:85 Chapter 15. 4 Waved Design Meteriel* 16A Tech Spec Boses pay, g 3,7 32

l ' System 80 +

Deslan crnerot Document CCW

, B 3.7.7 I BASES

- \

BACKGROUND ' (continued) components. The non-essential headers and the RCP headers j isolate on a low-low surge tank level. j Makeup water to the CCWS is normally supplied by the Demineralized Water System. The backup makeup water source

is from the Station Service Water System (SSWS). {

t

) The CCWS serves as an intermediate cooling water system i i between the Reactor Coolant System (RCS) and the SSWS. A : radiation monitor is provided at the outlet of the component 4 cooling water pumps to detect any radioactive leakage into the CCMS. 1 Additional information on the design and operation of the a system, along with a list of components served, can be found in EEEENFGD Chapter 9 (Ref. 1). z Q-Qb APPLICABLE The CCWS, in conjunction with the Station Service Water SAFETY ANALYSES System (SSWS) and the Ultimate Heat Sink (VHS), is capable of removing sufficient heat from the essential heat exchangers to ensure a safe reactor shutdown and cooling following a postulated accident coincident with a loss of offsite power. The CCWS, in conjunction with the SSWS, is capable of maintaining the outlet temperature of the component cooling water heat exchanger within the limits of 65* and 120*F during a design basis accident with loss of offsite power. A single failure of any component in the CCWS will not impair the ability of the CCWS to meet its functional requirements. The CCWS, in conjunction with the SCS and SSWS, is designed to cool the reactor coolant from 350*F to 140*F through the l shutdown cooling heat exchangers and the component cooling , water heat exchangers. The reactor can be cooled to 140*F 1 within 24 hours after reactor shutdown by first cooling the reactor coolant to 350*F through the steam generators and then cooling to 140*F by utilizing both divisions of SCS, ) CCWS, and SSWS. l (continued) , AsnerondDesign Materiel 16A Tech Spec Bases Page B 3.7-34 4

Syntem 80+ Design ControlDocument CCW B 3.7.7 BASES i SURVEILLANCE SR 3.7.7.2 (continued) A REQUIREMENTS W% pass the Surveillance when performed at the (18] month 4 Frequency. Therefore, the Frequency is acceptable from a reliability standpoint. h SR 3.7.7.3 ,, This SR verifies proper automatic operation of the CCW pum; = '=/ on an actual or simulated actuation signal. The CCW Systei is a normally operating system that cannot be fully actuat-as part of routine testing during normal operation. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. Operating experience has shown these components usually pass the Surveillance when performed at the [18] month Frequency. Therefore, the Frequency is acceptable from a reliability standpoint. REFERENCES 1. N Chapter 9. % vondDes&n Material 16A Tech spec Bases page g 3,7 3g

1

  . Svatem 80 +                                                       Deslas controlDocument 1                                                                                                 SSWS

{ B 3.7.8 l B 3.7 PLANT SYSTEMS fB3.7.8 Station Service Water System (SSWS) _ ,

BASES l

i BACKGROUND The Station Service Water System (SSWS) provides a heat sink i for the removal of process and operating heat from safety related components during a transient or DBA. During normal 4 operation or a normal shutdown, the SSWS also provides this function for various safety related and non-safety related components through the CCWS. i The SSWS consists of two separate, redundant, open loop, i safety related divisions. Each division cools one of two i divisions of the CCWS, which in turn cools 100% of the safety-related loads. The SSWS operates at a lower pressure t than the CCWS to prevent contamination of the CCWS with raw ! water. i Each division of the SSWS :onsists of two pumps, two strainers, two sump pumps, and associated piping, valves, controls and instrumentation. The station service water j pumps circulate cooling water to the component cooling water i heat exchanger and back to the ultimate heat sink. Provisions are made to ensure a continuous flow of cooling water under normal and accident conditions. i i Additional information about the design and operation of

p f, k SSWS, along with a list of the components served, can be l

found in 1:8848#*RP Chapter 9 (Ref. 1). a -

l C, E.CS AR-ce, ! APPLICABLE The SSWS, in conjunction with the Component Cooling Water SAFETY ANALYSES System (CCWS) and Ultimate Heat Sink (UHS), is capable of removing sufficient heat to ensure a safe reactor shutdown ] coincident with a loss of offsite power. j The SSWS is capable of maintaining the.CCWS supply temperature of 120*F or less following the design basis : accident under the most adverse historical meteorological l j conditioris consistent with the intent of Regulatory Guide j 1.27. ]! (continued) i y o=+,uwwsur~ssens~~ r s .m, I

         -                    -- .                                                                                             .\

Syntem 80+ Deskn ControlDocument SSWS ! B 3.7.8 ' 1 BASES (continued) SURVEILLANCE SR 3.7.8.1 '1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and motor operated, valves in the SSWS flowpath provides assurance that the proper flowpaths exist for SSWS operation. This SR does not apply to valves which are locked, sealed, or otherwise secured in position, since they were verified to be in the correct position prior to ~ locking, sealing, or securing. This SR also does not apply to valves which cannot be inadvertently misaligned, such as check valves. Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR is modified by a Note indicating that the isolation of the SSWS components or systems may render those components inoperable but does not affect the - OPERABILITY of the SSWS. The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions. SR 3.7.8.2 The SR verifies proper automatic operation of the SSWS pumps on an actual or simulated actuation signal. The SSWS is a normally operating system that cannot be fully actuated as part of the normal testing during normal operation. The [18] month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. - Operating experience has shown that these components usually pass the Surveillance when performed at the [18] month Frequency. Therefore, the Frequency is acceptable from reliability standpoint. REFERENCES 1. CEEEE UB Chapter 9. 4 - Annroved Design MeterW 16A Tech Spee Banee 9 g 3,742

T i Syntem 80+ N&n ControlDocument UHS B 3.7.9 i B 3.7 PLANT SYSTEMS i { B 3.7.9 Ultimate Heat Sink (VHS) , ,l t ) BASES { l BACKGROUND The UHS provides a heat sink for process and operating heat i from safety related components during a Design Basis l Accident (DBA) or transient, as well as during normal

operation. This is done utilizing the Station Service Water "

System (SSWS) and the Component Cooling Water (CCWS) System. : i 1 l The UHS has been defined as that complex of water sources I including necessary retaining structures (e.g., a pond wi-i its dam, or a river with its dam), and the canals or D e' k l . conduits connecting the sources with, but not including, - l cooling water system intake structures, as discussed in ; ! ftStatzth, Chapter 9 (Ref.1). .If cooling towers or - portions thereof are required to accomplish the UHS safet.. ) functions, they should meet the same requirements as the 05 54 # #~ g > ! sink. The two principal functions of the UHS are the l dissipation of residual heat after reactor shutdown, and j dissipation of residual heat after an accident. 1 A variety of complexes are used to meet the requirements for a UHS. A lake or an ocean may qualify as a single source. ' 4- If the complex includes a water source contained by a : structure, it is likely that a second source will be l required. ' j The basic performance requirements are that a 30 day supply of water be available, and that the design basis temperatures of safety related equipment not be exceeded. ! Basins of cooling towers generally include less than a 30 l

day supply of water, typically 7 days or less. A 30 day (

supply would be dependent on another source (s) and a makeup l system (s) for replenishing the source in the cooling tower basin. For smaller basin sources, which may be as small as a 1 day supply, the s the backup source (s) ystems become of for replenishing sufficient the basin importance that and the makeup system itself may be required to meet the same i design criteria as an Engineered Safety Feature (e.g., single failure considerations, and multiple makeup water sources may be required). (continued) Approwd Du&r niatwW 16A Tuh Son Buu Page B 3.7 43

1 ~ Syst m 80+ Deelan Control Document i s UHS j 5 3.7.9 1 j BASES 4 i j BACKGROUND It follows that the many variations in the VHS ! (continued) configurations will result in many unit-to-unit variations i in OPERABILITY determinations and SRs. The ACTIONS and SRs l } are illustrative of a cooling tower UHS without a makeup j j requirement. J \ l j Additional information on the design and operation of the j system, along with a list .of components served, can be found , l j in Reference 1. ) APPLICABLE The UHS removes heat from the reactor core following all l SAFETY ANALYSES accidents and Anticipated Operational Occurrences (A00s) in j which the plant is cooled down and placed on shutdown

;                          cooling. For those plants using it as the normal heat sink j                            for condenser cooling via the Condenser Circulating Water j                           System, plant operation at full power is its maximum heat                                  -

load. Its maximum post-accident heat load occurs after a design basis Loss Of Coolant Accident (LOCA). At this time, 3 recirculation of the IRWST through the Containment Spray z System and Safety Injection System is required to remove the core decay heat. i The operating limits are based on a conservative heat j transfer analyses for the worse case LOCA. Refer to'tBSeth C i y Chapter 9 (Ref.1) for details of the assumptions used in b ' g, the analysis. These assumptions include: worst expected meteorological conditions, conservative uncertainties when

calculating decay heat, and the worst case single active i failure. The UHS is designed in accordance with Regulatory l Guide 1.27 (Ref. 2) which requires a 30 day ~ supply of
cooling water in the UHS.

} The Ultimate Heat Sink satisfies the requirements of . '] Criterion 3 of the NRC Policy Statement. 1 i LCO The UHS is required to be OPERABLE. The UHS is considered i ! OPERABLE if it contains a sufficient volume of water at or j below the maximum temperature that would allow the SSWS to

;                          operate for at least 30 days following the design basis LOCA 5

without the loss of NPSH and without exceeding the maximum j design temperature of the equipment served by the SSWS. To } I l (continued) j i

    .w :Daten neuterW 16A ruh spec suu i                                                                                          Pero a 2.7 44 e

i

                                                                     . - -          L_                  .,
                                                                                                                         ..,I

Syotem 80 + Deslan control Document - UHS 8 3.7.9 BASES

r l SURVEILLANCE SR 3.7.9.3 (continued) - A1 l REQUIREMENTS l corrective action. The 31 day Frequency is based on l operating experience, the known reliability of the fan units, the redundancy available, and the low probability of l significant degradation of the VHS cooling tower fans occurring between surveillances.] l REFERENCES 1. N Chapter 9.

2. Regulatory Guide 1.27 (Rev. 01), " Ultimate Heat Sink for Nuclear Power Plants."

l l Approved Desker Material 16A Tech Spec Bases Pope B 3.7 47

Svatem 80+ Deslan ControlDocument Fuel Storage Pool Water Level B 3.7.10 B 3.7 PLANT SYSTEMS l B 3.7.10 Fuel Storage Pool Water Level BASES BACKGROUND The minimum water level in the Fuel Storage ~ Pool meets the assumptions of Iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are at their maximum capacity. The water also provides shielding during the movement of spent fuel. MbG. kC> 4. A general description of the Fuel Storage Pool design is e found in N Chapter 9 (Ref. 1). The assumptions of the fuel handling accident are found in 1:E332824 Chapter 15 4 - (Ref. 2). APPLICABLE The minimum water level in the Fuel Storage Pool meets the SAFETY ANALYSES assumptions of the fuel handling accident described in RG 1.25 (Ref. 3). The resultant two hour thyroid dose to a person at the exclusion area boundary (EAB) is well within the 10 CFR 100 (Ref. 4) limits. I According to Reference 3, there is 23 feet of water between the top of the damaged fuel bundle and the fuel pool surface for a fuel handling accident. With 23 feet water level, the assumptions of Reference 3 can be used directly. In , practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single ' bundle, dropped and lying horizontally on top of the spent fuel racks, however, there may be less than 23 feet above the top of the fuel bundle and the surface by the width of I the bundle. To offset this small non-conservatism, the analysis assumes that all 236 fuel rods fail, although i analysis shows that only the first four rows, 60 fuel rods, fail from a hypothetical maximum drop. The fuel storage pool water level satisfies Criterion 3 of the NRC Policy Statement. 4 (continued) AmendDu> AteewW.16A Tuh Spec Buu page g 3,7 4g

System 80+ Design Control Document l Fuel Storage Pool Water Level B 3.7.10 BASES SURVEILLANCE SR 3.7.10.1 (continued) - ' REQUIREMENTS the volume in the pool is normally stable. Water. level changes are controlled by plant procedures and are acceptable, based on operating experience. During refueling operations, the level in the fuel storage pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with LCO 3.9.6, " Refueling Water Level." REFERENCES 1. ws ry Chapter 9. A

2. _CE55?;E' Chapter 15. %

3. Regulatory Guide 1.25 (Rev. 00), " Assumptions Used fo-Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors."

4. 10 CFR 100, Reactor Site Criteria.

i Rpprived Design Material .16A Tech Spec Bases Page B 3.7-50

i System 80+ t=' n conend Document ADVs 8 3.7.11

B3.7 PLANT SYSTEMS i

8 3.7.11 Atmospheric Dump Valves (ADVs) l -

BASES BACKGROUND The Atmospheric Dump Valves (ADVs) provide a safety grade method for cooling the plant to Shutdown Cooling System (SCS) entry conditions should the preferred heat sink via the Steam Bypass System to the condenser not be available.

This is done in con, junction with the Emergency Feedwater 3 System providing cooling water from the Emergency Feedwater Storage Tank (EFWST). The ADVs may also be required to meet 7 the design cooldown rate during a normal cooldown when steam pressure drops too low for maintenance of a vacuum in the condenser to permit use of the Steam Bypass System. Four ADV lines are provided. Each ADV line consists of one l ADV and an associated block valve. Two ADV lines per steam generator are required to meet single failure assumptions - 2 ' following an event rendering one steam generator unavailable for Reactor Coolant System (RCS) heat removal. The ADVs are provided with upstream block valves to permit their being tested at power and to provide an alternate means of isolation. The ADVs are electrically operated with internal solenoid operated pilot valves and electronic valve positioning circuits to permit control of the cooldown rate. -

The ADVs are OPERABLE with only a DC power source available.

i A description of the ADVs is found in 3EEEEEEGft. Chapter 10. T - l (Ref. 1). g, The design basis of the Atmospheric Dump Valves (ADVs) is TPPLICABLE established by the the capability to cool the plant to 1AFETY ANALYSES i shutdown cooling entry conditions at the design rate of l [100'F/hr] using both steam generators, each with two ADVs. i This design is adequate to cool the plant to SCS entry ; conditions with only one ADV and one steam generator utilizing the cooling water supply available in the EFWST. i

                                                                                                     '(continued) onerovenf Doelps Ateeeniel 16A Tech Spec Basen page a 2.7 51

Svitem 80+ oeskn control Document 2 ADVs B 3.7.11 ) BASES

jAPPLICABLE in the accident analysis presented in Chapters 6 and 15 at f- M
SAFETY ANALYSES W , the ADVs are not assumed to be used until the 1 (continued) operator takes action to cool down the plant to SCS entry conditions for accidents accompanied by a loss of offsite power. Prior to the operator action, the main steam safety

valves (MSSVs) are used to maintain the steam generators

! pressure and temperature at the MSSVs setpoint. This is typically 30 minutes following initiation of an eve.nt. i (This may be less for a Steam Generator Tube Rupture (SGTR) - event). The limiting events are those which render one j steam generator unavailable for RCS heat removal, with a i coincident loss of offsite power as a result of turbine trip i and the single failure of one ADV on the unaffected steam

generator. Typical initiating events falling into this category are a main steam line break (MLSB) upstream of the 1

main steam isolation valves, a feedwater line break (FWLB), i and a SGTR event (although the ADVs on the affected steam 4 generator may still be available following a SGTR event). - t i The design must accommodate the single failure of one ADV to open on demand; thus, each steam generator must have at least two ADVs. The ADVs are equipped with block valves in j the event an ADV spuriously opens, or fails to close during g use. ] The ADVs satisfy Criterion 3 of the NRC Policy Statement. j. i LC0 [Two] ADV lines are required on each steam generator to ensure that at least one ADV is OPERABLE to conduct a plant cooldown following an event in which one steam generator becomes unavailable, accompanied by a single active failure of one ADV line on the unaffected steam generator. The block valves must be OPERABLE to isolate a failed open ADV. A closed block valve does not render it or its ADV line inoperable if operator action time to open the block valve is supported in the accident analysis.

Failure to meet the LC0 can result in the inability to cool ! the plant to SCS entry conditions following an event in j which the condenser is unavailable for use with the Steam j Bypass System. > I 4 (continued) ! someou+,mromt 1sa rech see sua r,o. s 2.7.s2

1 Syntem 80+ Design ControlDocument ' l ADVs B 3.7.11 BASES ACTIONS C.l. C.2. and C.3 (continued) - [12] hours; and MODE 5 in [24] hours. The allowed Completion Times are reasonable based on operating experience to reach the required MODES from full power operation without challenging plant systems. SURVEILLANCE SR 3.7.11.1 REQUIREMENTS To perform a controlled cooldown of the RCS, the ADVs must be able to be opened and throttled through their full range. This SR ensures the ADVs are tested through a full control cycle at least once per fuel cycle. Performance of inservice testing, or use of an ADV during a plant cooldown may satisfy this requirement. This Surveillance Frequency is based on the length of a fuel cycle and has been shown to be adequate through operating experience. SR 3.7.11.2 The function of the block valve is to isolate a failed open ADV. Cycling the block valve closed and open demonstrates its capability to perform this function. Performance of inservice testing, or use of the block valve during plant cooldown may satisfy this requirement. The Surveillance interval of 18 months is based on engineering judgment, and : has been shown to be acceptable through operating i experience. l l REFERENCES 1. CE32temeG Chapter 10. y l l l 1 1 l 4 Approved Desly Material 16A Tech Spec Bases Pop B 3.7 5,4 1

l i Sv' tem 80+ Deslan ControlDocument !j CCVS e B 3.7.12 I i BASES 4 3 The Technical Support Center (TSC) air-conditioning system ) BACKGROUND 1 (continued) consists of an air-handling unit, return air and sitoke purge J fans, and an emergency filter unit. The computer room air-

!                       conditioning system consists of two 100% air-conditioning

units and associated fans. Both the Technical Support

} Center and computer room air-handling systems are non-safety ) and non-seismic.

I The balance of CCVS consists of two redundant air-handling l units, each with roughing filters, essential chilled water 1 cooling coils and fans serving Division I electrical rooms, channel A and channel C. Two equal units are serving I Division II channel B and D. Each division will function

!                       with one of the redundant air handling units delivering

filtered, conditioned air to the various electrical i equipment rooms, including essential battery rooms. Chilled i water is supplied from the Essential Chilled Water System.

i Each division also contains redundant battery rooms with - ! fans operating continuously to maintain the hydrogen ! concentration below two percent. Outlet ducts in battery i rooms are located near the ceiling for hydrogen control. j The safe shutdown area is sarved by Division II. 1 i Return air from the various essential electrical equipment ! areas is mixed with a portion of outside air for

!                      ventilation, is filtered and conditioned in the air-handling i                       unit, and is delivered to the rooms through supply ductwork.

. Duct-mounted heating coils provide final adjustments to l temperature in selected equipment rooms. i i The Operation Support Center (OSC), Personnel Decon Rooms, Break Room, Shift Assembly and Offices, and Radiation Access Control areas all are served by an individual air handling unit consisting of a centrifugal fan, non essential chilled j water coil and roughing filter. The MCR and TSC receive outside air from the cleanest of two

sources.

i~

!                      A single division will pressurize the MCR to at least                                       !

3 [0.125] inches water gauge, and provides an air l recirculation rate in excess of [25%] per hour. The Centrol Complex Ventilation Systems operation in maintaicing the MCR

habitable is discussed in CSSSAR*36. Chapter 6 (Ref.1). ~

f 1- D O i l l j (continued) I 1

  %. A De ipn Mewiel 16A Tuh Spec Buu                                         Pope B 3.7 56 k                                                                                                                    l 4

System 80+ Desian controlDocument CCVS B 3.7.12 l ! BASES l - . BACKGROUND Redundant supply and recirculation divisions provide the - -

(continued) required filtration should an excessive pressure drop develop across the other filter division. Normally open isolation dampers are arranged in series pairs so that the :

failure of one damper to shut will not result in a breach of isolation. Redundant detectors for radiation and toxic gas protection are provided. The Control Complex Ventilation System is designed in accordance with Seismic Category I _ requirements.

a The Control Complex Ventilation System is designed to j maintain the MCR environment for 30 days continuous j occupancy after a DBA without exceeding 5 rem whole body dose. q The air entering the MCR is continuously monitored by radiation and toxic gas detectors. One detector above the setpoint will cause actuation of the emergency radiation - mode or toxic gas isolation mode as required. ) APPLICABLE The Control Complex Ventilation System components are SAFETY. ANALYSES arranged in redundant safety-related ventilation divisions. ]

'                                 The location of components and ducting within the control room envelope ensures an adequate supply of filtered air to all areas requiring access. During emergency operation the 1

Control Complex Ventilation System maintains the MCR temperature between 73*F and 78'F. The Control Complex Ventilation System provides airborne radiological protection for the control room operators as demonstrated by the control room accident dose analyses for the most limiting d . D d. ) design basis Loss Of Coolant Accident fission product release presented in 4GEElutiWe. Chapter 15 (Ref. 2). t ) The analysis of toxic gas releases demonstrates that the toxicity limits are not exceeded in the control room following a toxic chemical release. i i The worst case single active failure of a component of the , { Control Complex Ventilation System, assuming a loss of l l offsite power, does not impair the ability of the system to perform its design function. 1 l s (continued)

  '                                                                                                                    l AtsprovedDeslers Meteriel.16A Tech Spec Bases                                      Page B 3.7 57

j Svatem 80+ Deslan Conuel Document CCVS

B 3.7.12 BASES (continued)

SURVEILLANCE SR 3.7.12.3 - - REQUIREMENTS i (continued) This SR verifies each CCVS division starts and operates on en actual or simulated actuation signal. The Frequency of (18] months is consistent with that specified in Reference , 3.

SR 3.7.12.4 4 This SR verifies the integrity of the control room enclosure and the assumed inleakage rates of potentially contaminated air. The control room positive pressure, with respect to , potentially contaminated adjacent areas, is periodically I tested to verify proper tunction of the CCVS. The CCVS is designed to pressurize the control room 2 [0,125] inches water gauge positive pressure with respect to adjacent areas in order to prevent unfiltered inleakage. The CCVS is * , designed to maintain this positive _ pressure during the J [ pressurization]. mode of operation at a rate of [s 2000] cfa. The Frequency of [18] months on a STAGGERED TEST BASIS

is consistent with the guidance provided in NUREG-0800,

;                         Section 6.4 (Ref. 4).

REFERENCES 1. :Cgsame.ne Chapter 6. c

2. "tSSSWW9e Chapter 15. 4

3. Regulatory Guide 1.52 (Rev. 02), " Design, Testing an Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Coole Nuclear Power Plants."

j 4. NUREG-0800, " Standard Review Plan," Section 6.4,

                                  " Control Room Habitability System," Rev. 2, July 1981.

l l l

4 Anwoved Deekn AtewW 16A Tech Spec Buee Pepe B 3.7-61

              , .-                      ,                 n    -                             - - , ,     -

System 80+ i Deslan ControlDocumnt CRVS B 3.7.13 B 3.7 PLANT SYSTEMS B 3.7.13 Control Room Ventilation System (CRVS) _ , BASES BACKGROUND The CRVS provides temperature control for the Main Control Room (MCR) following isolation of the MCR. The CRVS consists of two independent, redundant divisions ~

,                                which provide cooling and heating of recirculated MCR air.

Each division consists of a heating coils, cooling coils, instrumentation and controls to provide for MCR temperature control. The CRVS is a sub-system providing air temperature control for the " Control Complex Ventilation System," LCO 3.7.12.

The CRVS is a standby system, parts of which may also

 ;                               operate during normal plant operations. A single division will provide the required temperature control to maintain the MCR between 73*F and 78'F. The CRVS operation in maintaining the MCR temperature is discussed in SEGENR:90:. t Chapter 9 (Ref. 2).

d DO The design basis of the CRVS is to maintain the MCR APPLICABLE environment throughout 30 days continuous occupancy. The SAFETY ANALYSES CRVS operation in maintaining the MCR temperature is y discussed in liEEERBliin Chapter 9 (Ref. 2). The CRVS components are arranged in redundant safety related , divisions. During emergency operation, the CRVS maintains-the temperature between 73*F and 78'F. A single active

failure of a component of the CRVS, assuming a loss of offsite power, does not impair the ability of the system to perform its design function. Redundant detectors and controls are provided for control room temperature control.

1 CRVS is designed in accordance with Seismic Category I requirements. CRVS is capable of removing sensible and latent heat loads from the control room which include consideration of equipment heat loads and personnel

,                               occupancy requirements to ensure equipment OPERABILITY.

The CRVS satisfies Criterion 3 of the NRC Policy Statement.

1 (continued)

 ! Anweved Deelers nieserial 16A Tech Spec Bones                                       Page B .1.7 62

System 80+ c:b= coned Document CRBS B 3.7.13 I BASES (continuad) SURVEILLANCE SR 3.7.13.1 - REQUIRENENTS l This SR verifies that the heat removal capability of the l system is sufficient to meet design requirements. This SR consists of a combination of testing and calculations. An (18] month Frequency is appropriate, since significant degradation of the CRVS is slow and is not expected over this time period.

REFERENCES 1. - M Chapter 6.

2. N Chapter 9.

3. Regulatory Guide 1.52 (Rev. 2).

4. NUREG-0800, " Standard Review Plan," Section 6.4,

                         " Control Room Habitability System," Rev. 2, July 1981.

en W W **"W 1s^ run sm nun w s 2.1.ss

___ _ ___._ _ _ _ _ _ _ _ _ _ . . . _ . ~ _ __ _ _ _ _ _ .. ___ _ i i System 80+ Mn controlDocument j SBVS

B 3.7.14

        .B 3.7 PLANT SYSTEMS j          B 3.7.14 Subsphere Building Ventilation System (SBVS)

. i

~ BASES i BACKGROUND The SBVS consists of a general supply and exhaust ventilation system that performs heat removal and air ! exchange functions. The ventilation system is supplemented

!                                   by individual cooling units and ventilation fans that serve essential mechanical equipment areas. The SBVS serves all                                         ~~

j areas of the subsphere. 1 The SBVSs are separated according to divisions, with each

100% exhkust system containing a filter train consisting of

}                                  a moisture eliminator, prefilter, electric heater, absolute (HEPA) filter, carbon adsorber, post filter (HEPA), along with ducts and valves, related instrumentation and two 100%

j capacity fans. 1 The SBVS is discussed in N Chapters 9 and 15 d ! (Refs. I and 2 respectively), as it is used for normal, as G DO j well.as post accident, atmospheric cleanup functions. The primary purpose of the heaters is to maintain the relative humidity at an acceptable level consistent with iodine removal efficiencies, as discussed in the Regulatory Guide 1.52 (Ref. 3). (Carbon adsorbers are not credited for ) iodine removal in SBVS.) ' The design basis of the SBVS is established by the large APPLICABLE break LOCA. The system evaluation assumes a passive failure SAFETY ANALYSES of the SIS outside containment, such as safety injection i pump seal failure. In such a case, the system limits the radioactive release to within 10 CFR 100 limits (Ref. 4). The analysis of the effects and consequences of a large break LOCA is presented in Reference 2. The SBVS satisfies Criterion 3 of the NRC Policy Statement. 1 I j LCO , Two independent and redundant divisions of the Subsphere '

!                                 Building Ventilation System are required to ensure that at j                                  least one is available, assuming a single failure coincident I.

i (continued) { l Anrendou+r nwerw. su rech so,e sua r.,, a s.7.ee a

System 80+ oeskn contrat Document SBVS B 3.7.14 BASES (continued) REFEnENCES 1. 'C555AR:GGb. Chapter 9. *

2. 6 Chapter 15.

3. Regulatory Guide 1.52 (Rev. 02), " Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmospheric Cleanup System Air Filtration and Adsorption Units of Light-Water Cooled ~

Nuclear Power Plants."

4. 10 CFR 100, Reactor Site Criteria.

5. NUREG-0800, " Standard Review Plan", Section 6.5.1, Rev. 2, "ESF Atmosphere Cleanup Systems", Rev. 2, July 1981.

l l { Anwend Design Materiel 16A Tech Spec Bases y,y, y 3, y,79 l l

I Svstem 80+ Deskrn ConeelDocenent FBVES B 3.7.15 l B 3.7 PLANT SYSTEMS j , B 3.7.15 Fuel Building Ventilation Exhaust System (FBVES)

                                                                                                                                                              *1 BASES                                                                                                                                             I BACKGROUND                       The FBVES filters airborne radioactive particulates from the area of the fuel pool following. a fuel' handling accident.

The FBVES, in conjunction with other, normally operating i systems also provides environmental control of temperature

and humidity in the fuel pool area.

I The Fuel Building Ventilation Exhaust System consists of two independent, redundant divisions. Each of the divisions I consists of a moisture eliminator, prefilter, electric heater, absolute filter, and post filter. It is equipped with a bypass section. The normal mode of operation for the filter trains is in the bypass position. Radiation detection is provided in the duct system header, upstream of the filter train inlet to monitor radioactivity. Upon indication of high radioactivity in the exhaust duct system, the bypass dampers will automatically close and the filter train inlet dampers will automatically open to direct air flow through the filter trains. Air from the Fuel Building 1 Exhaust System is directed to the unit vent, where it is monitored before release to the atmosphere. The exhaust air system is manually set to the filtered mode i during all fuel handling operations. S p6 The FBVES is discussed in Chapter 9 N (Ref.1). 1 It may be used for normal, as well as post-accident atmospheric cleanup functions. APPLICABLE The FBVES is designed to mitigate the consequences of a fue' SAFETY ANALYSES handling accident in which [all] rods in the fuel assembly are' assumed to be damaged. The analysis of the fuel handling accident is given in Reference 2. The Design Basis Accident analysis of the fuel handling accident assumes that ' only one division of the FBVES functional, due to a single failure that disables the other division. The accident analysis accounts for the reduction in airborne radioactive material provided by the remaining one division of this filtration system. The amount of fission products available (continued) l L.= ouw moeuw. su rus seu suu p ,, , 3. 7. , , ;

System 80+ oesion controlDocument 't FBVES B 3.7.15 a 1 4 ] BASES (continued) 1 4 - 4 REFERENCES 1. N Chapter 9.

2. -rzman=sP Chapter 15. 4

;                            3.      Regulatory Guide 1.52 (Rev. 02), " Design, Testing and l

Maintenance Criteria for Post Accident Engineered-i Safety-Feature Atmospheric Cleanup System Air i Filtration and Adsorption Units of Light-Water Cooled _ l Nuclear Power Plants." i ! 4. Regulatory Guide 1.25, " Assumptions Used for { Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and

Storage Facility for Boiling and Pressurized Water: Reactors".

i 5. 10 CFR 100 - Reactor Site Criteria. , l 6. NUREG-0800, " Standard Review Plan", Section 6.5.1, Rev. 2, "ESF Atmosphere Cleanup Systems", July 1981. i i i l i i l 1 h i f ] I l j i I i ApprovedDesign Materiel 16A Tech Spec Boses page g 3,7 75

l Syntem 80+ Desian controlDocument DBVS B 3.7.16 1 BASES SURVEILLANCE SR 3.7.16.2 -

;      REQUIREMENTS

{ (continued) Periodically, the exhaust fans and their controls must be tested to verify proper performance. A Frequency of 18 months is judged to be adequate and is consistent with t'~

!                              major surveillance testing performed on the diesel j

J generators. REFERENCES 1. u n M Chapter 9. 4 I 1 1 4 E a i j I 1 i j i i l i i l { 4prendDeskru Meterte! 16A Tech Spec Besos pay, y 3,7 79

System 80+ oeslan controloccument ECWS B 3.7.17 8 3.7 PLANT SYSTEMS B 3.7.17 Essential Chilled Water System (ECWS)

                                                                                                    'l 4 l BASES BACKGROUND                   The ECWS provides a heat sink for the removal of process and operating heat from selected safety related air handling systems during a Design Basis Accident (DBA) or transient.             !

I The ECWS is a closed loop system consisting of two ; independent divisions. Each 100% capacity division includes a chiller, heat exchanger, expansion tank, two pumps, chemical addition tank, piping, valves, controls, and instrumentation. An independent, 100% capacity chiller cools each division. The essential chiller is actuated on high ECWS temperature and supplies chilled water to essential HVAC units during a design basis event. During normal operation, the Normal Chilled Water System (NCWS) performs the cooling function of the ECWS through the ECWS heat exchanger with one of the ECWS pumps recirculating chilled water through the system. The NCWS is a non-safety related system. Additional information about the design and operation of the system, can be found in 1535195er. Chapter 9 4 (Ref. 1). APPLICABLE The design basis of the ECWS is to remove the post-accident SAFETY ANALYSES heat load from ESF spaces following a design basis accident with a loss of offsite power. Each division provides chilled water to the HVAC units at the design temperature of 45'F. The maximum heat load in the ESF pump room area occurs following a loss of coolant accident (LOCA). Hot fluid from the IRWST is supplied to the Safety Injection (SI) and Containment Spray (CS) Pumps. This heat load to the area atmosphere must be removed by the ECW System to ensure these systems remain OPERABLE. During a normal cooldown, the Shutdown Cooling System (SCS) piping also provides a heat load in areas served the by ECWS. The ECWS satisfies Criterion 3 of the NRC Policy Statement. (continued) Approved Des} prs Material- 16A Tech Spec Bases page g 3.7.go

i i Svatem 80+ Deslan ControlDocument I ECWS 8 3.7.17 i, BASES l 1 . , ACTIONS B.1 and B.2 (continued) ! must be placed in at least MODE 3 within 6 hours, and in { MODE 5 within 36 hours. The allowed Completion Times are i

reasonable, based on operating experience, to reach the

! required unit conditions from full power conditions in an j orderly manner without challenging unit systems.

SURVEILLANCE SR 3.7.17.1 REQUIREMENTS j Verifying the correct alignment for manual, power operated, j and automatic valves in the ECWS flowpath provides assurance i

that the proper flowpaths exist for ECWS operation. This SR i does not apply to valves which are locked, sealed, or ! otherwise secured in position, since they were verified to be in the correct position prior to locking, sealing, or {

securing. This SR also does not apply to valves which

! cannot be inadvertently misaligned, such as check valves.

This Surveillance does not require any testing or valve

! manipulation; rather, it involves verification that those ^ valves capable of potentially being mispositioned are in the correct position. The 31 day Frequency is based on engineering judgment, is i consistent with the procedural controls governing valve operation, and ensures correct valve positions. ' 4 SR 3.7.17.2 - 1 i This SR demonstrates proper automatic operation of the ECWS. The [18] month Frequency is based on the need to perform

this Surveillance under the conditions that apply during a l unit outage and the potential for an unplanned transient if l the Surveillance were performed with the reactor at power.

4 The (18] month Frequency is based on operating experience

and design reliability of the equipment.

. REFERENCES 1. 1:Essalliint, Chapter 9. C Anvoveet Design Aceterial.16A Tech M Resen Page E 3.7 82 i

Syntem 80+ Design ControlDocument MAIN STEAM LINE LEAKAGE B 3.7.18 B 3.7 PLANT SYSTEMS B 3.7.18 MAIN STEAM LINE LEAKAGE 1 l BASES BACKGROUND A limit on leakage from the main steam lines inside containment is required to limit system operation in the presence of excessive leakage. Leakage is limited to an ' amount which would not compromise safety consistent with the ' Leak-Before-Break (LBB) analysis discussed in Chapter 3.ar 'l

                              - (Reference 1). This leakage limit ensures                          d-appropriate action can be taken before the integrity of the            1    DO main steam lines is impaired.

LBB is an argument which allows elimination of design for dynamic load effects of postulated pipe breaks. The fundamental premise of LBB is that the materials used in nuclear plant piping are strong enough that even a large throughwall crack leaking well in excess of rates detectable by present leak detection systems would remain stable, and would not result in a double-ended guillotine break under maximum loading conditions. The benefit of LBB is the elimination of pipe whip restraints, jet impingement effects, subcompartment pressurization, and internal system blowdown loads.

                                                                                                         .Dd As described in Chapter 3 N (Reference 1),                          4     -

LBB has been applied to the four main steam line pipe runs inside containment. Hence, the potential safety significance of main steam line leaks inside containment t l requires detection and monitoring of any leakage inside containment. This LC0 protects the main steam lines inside containment against degradation, and helps assure that serious leaks or Main Steam Line Breaks (MSLBs) will not develop. The consequences of violating this LCO include the possibility of further degradation of the main steam lines, which may lead to an MSLB. APPLICABLE The safety significance of plant leakage inside containment SAFETY ANALYSES varies depending on its source, rate, and duration. Therefore, detection and monitoring of plant leakage inside ; containment are necessary. This is accomplished via the instrumentation required by LC0 3.4.14, "RCS Leakage (continued) Approved Design Materbel 16A Tech Spec Bases page a 3.7 83

Sy, tem 80+ Design ContrriDocument

MAIN STEAM LINE LEAKAGE i B 3.7.18 BASES APPLICABLE Detection Instrumentation," and the RCS water inventory -

SAFETY ANALYSES balance (SR 3.4.12.1). Subtracting RCS leakage as well as (continued) any other identified non-RCS LEAKAGE into the containment area from the total plant leakage inside containment

provides qualitative information to the operators regarding possible MAIN STEAM LINE LEAKAGE. This allows the operators to take corrective action should leakage occur which is detrimental to the safety of the facility and/or the public.

j MAIN STEAM LINE LEAXAGE inside containment is an initial assumption in the LBB analysis described in Chapter 3 et

                            -1:EEWWAIE.(Reference 1). As such, it satisfies the                    4 j                              requirements of Criterion 2 of the NRC Policy Statement.

i I LC0 MAIN STEAM LINE LEAKAGE is defined as leakage inside containment in any portion of the four (4) 28" I.D. main steam line pipe walls. Up to 10 gpm of MAIN STEAM LINE LEAKAGE is allowable because such leakage is within the . capability of the makeup system, and is well below the leak

rate calculated by the industry standard computer program 1

PICEP (Pipe Crack Evaluation Program, Reference 2) for the 1 LBB analyzed case of a main steam line crack twice as long l as a crack leaking at ten (10) times the detectable leak ] rate under normal operating load conditions. Violation of 1 this LC0 could result in continued degradation of the main steam lines. APPLICABILITY Because of elevated main steam system temperatures and pressures, the potential for MAIN STEAM LINE LEAXAGE is greatest in MODES 1, 2, 3, and 4. In MODES 5 and 6, a MAIN STEAM LINE LEAXAGE limit is not provided because the main steam system pressure is far 4 lower, resulting in lower stresses and a reduced potentici for leakage. In addition, the steam generators are not the ] primary method of RCS heat removal in MODES 5 and 6. 1 Approved Design Matedel 16A Tech Spec Bases (continued) Page B 17 84

!              Svstem 80+                                                                       w&n controloccanent MAIN STEAM LINE LEAKAGE j{                                                                                                             B 3.7.18 3

BASES (continued) 1 l ACTIONS A.1 and A.2 With MAIN STEAM LINE LEAKAGE in excess of the. LCO limit, the '! unit must be brought to lower pressure conditions to reduce { the severity of the leakage and its potential consequences. The reactor must be placed in MODE 3 within 6 hours and

!                                                   MODE 5 within 36 hours. This action reduces the main steam

} line pressure and leakage, and also reduces the factors ~ j which tend to degrade the main steam lines. The Completion - J Time of 6 hours to reach MODE 3 from full power without challenging plant systems is reasonable based on operating experience. Similarly, the Completion Time of 36 hours to 3 reach MODE 5 without challenging plant systems is also reasonable based on operating experience. In MODE 5, the

pressure stresses acting on the main steam lines are much lower, and further deterioration of the main steam lines is i less likely.

I 1 2 SURVEILLANCE SR 3.7.18.1 i REQUIREMENTS - j Verifying that MAIN STEAM LINE LEAKAGE is within the LCO

limit assures the integrity of the main steam lines inside t

containment is maintained. An early warning of MAIN STEAM i LINE LEAKAGE is provided by the automatic systems which j monitor the containment sump and containment cooler i condensate tank see LC0 3.4.14). MAIN STEAM LINE LEAKAGE i would appear as u(nidentified LEAKAGE inside containment via j these systems, and can only be positively identified by inspection. However, by performance of an RCS water inventory balance (SR 3.4.12.1) and evaluation of the i cooling and chilled water systems inside containment, i determination of whether the main steam lines are potential i sources of unidentified LEAKAGE inside containment is -- } possible. 1 [ REFERENCES 1. N Chapter 3. 4

2. PICEP, Pipe Crack Evaluation Program.

1 i 1 j AwevdDulon nietwW- 16A Tuh Son tuu page g 2,7.gg i

i System 80+ Deskn ControlDocument i AC Sources - Operating B 3.8.1 BASES l s BACKGROUND Certain required unit loads are returned to service in a

(continued) predetermined sequence in order to prevent overloading the DG in the process. Within [1] minute after the initiating
signal is received, all loads needed to recover the unit or maintain it in a safe condition are returned to service.

In accordance with Regulatory Guide 1.9 (Ref. 2), diesel

generators 1 and 2 have [6067] kW continuous and (6674] kW .

I two-hour load ratings. The diesel generators are rated at i (4160 volts), three phase, 60 Hz, and are capable of

attaining required frequency and voltage within twenty seconds after receipt of a start signal (Ref. 3). The ESF systems which are powered from divisional power sources are a listed in Reference 3.

d The CTG is a diverse and independent non-Class IE on-site power source provided for coping with a Station Blackout (580) and a Loss of Offsite Power (LOOP) scenarios. The CTG is located within the protected area and it will start l automatically, within [2] minutes from the onset of a LOOP , event. In addition, the CTG is automatically connected to the de-engergized 4.16 kV Permanent Non-Safety buses. Alignment to the class IE ESF buses is accomplished from the control room. The CTG is sized to accommodate one Safety Division loads for a worst case unit shutdown to cold shutdown and/or DBA and one division of Permanent Non-Safety loads. The CTG is Quality Class 2 and is designed with a High Confidence of Low Probability of Failure (HCLPF) value+ _ t ' W ""; "--- "-!::i: "...... J. .u ;n .t ' ~" M ua,t provides ass'.rance that the CTG will be available to back up the DGs for seismic events on the order of the design basis earthqune of .3g (Ref.15). This robust design includes

the enclosure and the support systems of the CTG. -

Other axternal events which could affect CTG availability as a backup to the DG are hurricanes and tornados. Due to early warning systems the plant will be required to shu own i as a hurricane approaches. For the tornado it is assume the CTG will not be available. (continued) Anwend ou5, ^***ww. saa run son nun rege s 3.s 3

Syntem 80+ Design ControlDocument AC Sources - Operating B 3.8.1 BASES l BACKGROUND A PRA for the CTG's contribution to core damage frequency (continued) (CDF) was performed.

                                          ,                     y Q ca;.; g                           s j .., d ; si...,
                               -- _ = : :   u_..._.  . . . . . . . . . . , . __
                                                                                                 -...m.                                       ,
                                     .-            -    .              . =            ts.iu v u 5uw                   us

d tr.: cic .;.: ;d . L r... ;... 1, n.; ::r :r. .. .u >

                              = = i ;t.:, i. ;; 2.;; :-0.

This PRA provides an assurance that the CTG can be substituted for the DG without adversely impacting CDF for internal events and tornado _ t strikes. h APPLICABLE The initial conditions of DBA and transient analyses in SAFETY ANALYSES m Chapters 6 (Ref. 4) and 15 (Ref. 5) assume ESF e D e/ Q systems are OPERABLE. The AC Power System is designed to g _ d (0 provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power tc ESF systems so that the fuel, Reactor Coolant System, and

containment design limits are not exceeded. These design limits are discussed in more detail in the Bases for LC0 Sections 3.2 (Power Distribution Limits), 3.4 (Reactor Coolant System), and 3.6 (Containment Systems). In general, the safety analysis considered offsite power to be available to ESF equipment following event initiation. Offsite power is not considered to be safety-related. A loss of offsite power (LOOP) alone is an analyzed event since it presents a challenge to the plant's safety features and would result in a total loss of AC power if the diesel generators failed to start. The OPERABILITY of an offsite AC source is not explicitly required by the safety analyses. Therefore, the need for two qualified circuits was not derived from the safety analysis, since events postulating failure of offsite power considered a complete loss of offsite power. Such events - disable all offsite circuits. The requirement for two qualified circuits was derived from the design criteria (Ref.1) and standards incorporated into the plant design, which required redundant, independent offsite power sources. 4 l l i (continued) I l AnnrevedDee&n Adatoriel- 16A Tech Spec Beses page g 3.g.4 i i _

> Syntem 80+ Deslan controlDocument AC Sources - Operating B 3.8.1

BASES APPLfCABLE The OPERABILITY of the AC electrical power sources is ' *

;    SAFETY ANALYSES             consistent with the initial assumptions of the accident (continued)             analyses and is based upon maintaining at least one division of the AC and DC Power Sources and associated distribution systems OPERABLE during accident conditions in the event of

(1) an assumed loss of all offsite or all ~onsite AC power, l and (2) a worse case single failure.

The AC sources satisfy Criterion 3 of the NRC Policy

Statement.

s l l LC0 Two qualified circuits (Ref. 3) between the offsite

!'                               transmission network and the onsite Class IE AC Distribution System, and the two independent diesel generators (Ref. 3) each capable of supplying one division of the onsite Class IE AC Distribution System, ensure availability of the required power to shutdown the reactor and maintain it in a safe shutdown condition after an anticipated operational occurrence (A00) or a postulated design basis accident (OBA).

j Qualified offsite circuits are those that are described in 'D b i i

M"-E and are part of the licensing basis for the unit. *---

l b'd / Each offsite circuit must be capable of maintaining required frequency and voltage, and accepting required loads during an accident, while connected to the ESF buses. i The two circuits from offsite are physically independent such that a single component fault (e.g., breaker trip) will r not cause both power sources to be lost to one or more [4160 volt) emergency buses. Thus, a physically independent circuit consists of one incoming line to the [230 kV) i' (Preferred Switchyard Interface I) switchyard, a circuit path (including breakers and disconnects) to one energized UAT (X or Y), and a circuit path from the energized UAT to the associated [4160 volt) emergency buses. A physically i independent circuit also consists of the incoming line to 4 the Preferred Switchyard Interface II, a circuit path (including breakers and disconnects) to the one energized RAT (Division I or II), and a circuit path from that energized RAT to its 4160 volt emergency buses. Each i l 1 (continued) Approved Desigrs Matedal- 16A Tech Spec Bases page g 3,s.$

1 I Sy~ tem 80 + Deslan C:ntrol Document i AC Sources - Operating ' B 3.8.1 BASES SURVEILLANCE Diesel Generator Test Schedule (continued) REQUIREMENTS prevented the diesel generator from performing its intended safety function. As such, they do not impact the . Surveillance Frequency of the diesel generator that failed.

REFERENCES 1. 10 CFR50, Appendix A, General Design Criteria 17,

                                                " Electric Power Systems."

2. Regulatory Guide 1.9, " Selection, Design, and Qualification of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants,"

Revision 3.

3. , Chapter 8. .

4. tMet#=fra Chapter 6. < _

5. 4GEaut93; Chapter 15. >

6. ANSI C84.1-1982.

7. ASME Boiler & Pressure Vessel Code Section XI.

8. Regulatory Guide 1.93, " Availability of Electric Power Sources," December 1974.

9. Generic Letter 84-15, " Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability,"

July 2, 1984.

10. 10 CFR 50, General Design Criteria 18, " Inspection and Testing of Electric Power Systems."

11. Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as On-site Electric Power Systems at Nuclear Power Plants," August 1977.

12. Regulatory Guide 1.137, " Fuel Oil Systems for Stn.dby ;

Diesel Generators," October 1979. l l (continued) Approvent Design Meteria! 16A Tech Spec Basee pay, g 3,g.37 1

Syntem 80+ Design ControlDocument AC Sources - Operating B 3.8.1 BASES REFERENCES 13. -ws9 Chapter 9. " ' (continued)

14. IEEE 308-1974, "IEEE Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations."

15. :6EiMERigg Chapter 19.

x e l l l Anvend Des &n Marwiel 16A Tech Spec Buu 9 g 3,g.3g

i i Sv' tem 80+ &&n CrntrolDocument AC Sources - Shutdown j B 3.8.2 J l BASES i j LCO In addition, proper sequencer operation is an integral part '

,     (continued)                   of offsite circuit OPERABILITY if its inoperability in any                                                         !
;                                  way impacts oa the ability to start and maintain energized

any loads required OPERABLE by LCO 3.8.10.

i Certain diesel generator support systems are addressed in other LCOs. During inoperabilities in these support

!                                   systems, inoperable diesel generators do not necessarily

result unless specifically directed by Required Actions '

l ! 7 j (refertoLCO3.0.M)5/984,, Appe. ' q As described i g" Shutdown Risk Evaluation 6 " (Ref. 1), " j in the event of an accident during shutdown, the TS are p 4 l# g , designed to maintain the plant in a condition such that, " g cl

l even with a single failure, the plant will not be in

immediate difficulty. a g, p-b, ,

I i i APPLICABILITY The MODE 5 and MODE 6 APPLICABILITY assures AC power sources , j . are OPERABLE to support the equipment required to be j OPERABLE for the various conditions of these MODES. i MODE 5 has a subset of plant operating conditions. These j include: 1) MODE 5 - Loops Filled, 2) MODE 5 - Loops Not i Filled, and 3) MODE 5 - REDUCED RCS INVENTORY. During these j different operating conditions the systems required to be j OPERABLE are dictated by the conditions. Identifying the 3 required systems and the Distribution Systems to support ! them is essential in determining the number of qualified j circuit (s) and the number of on-site source (s) that must be OPERABLE in these different conditions. MODE 6 also has a subset of plant operating co.,ditions. 1 These include: 1) MODE 6 - High Water Level, 2) MODE 6 - Low Water Level, and 3) MODE 6- REDUCED RCS INVENTORY. As 3 discussed in the previous paragraph, identification of the required systems for each condition must be accomplished to ensure compliance with the TS. The AC power sources that are required to be OPERABLE in i MODES 5 and 6 provides assurance that: i 4 l i

(continued) f 4pred Deefpn Atatoriel 16A Tech Spec Basee page g 2.g 4 i.

Sy' tem 80 + Design ControlDocument AC Sources - Shutdown , B 3.8.2 l BASES ACTIONS C.4. C.S. and C.6 (continued) ' i time for repairs, and the low probability of. a DBA occurring t during this period. In Condition C, if the CTG is available, the 14 day Completion Time of Required Action C.6 takes into account the ability of the CTG to automatically start and to be aligned to the ESF buses in [10] minutes. (See the ! Background section for a discussion of the CTG.) ! D.1. D.2.1. D.2.2. D.2.3. D.2.4. and D.2.5 If the Required Actions and Completion Times of Condition C are not met Condition D must be entered. The Required Actions of Condition D are the same as those of Condition B. A discussion of this bases can be found in the above paragraphs for those Required Actions of Condition B. SURVEILLANCE SR 3.8.2.1 REQUIREMENTS SR 3.8.2.1 requires the SRs from LC0 3.8.1 that are necessary for ensuring the OPERABILITY of the AC sources in other than MODES 1, 2, 3, and 4. This SR is modified by a Note. This Note requires the performance for SR 3.8.1.3 only when the number of AC sources is more than the minimum required by LCO 3.8.2. This precludes requiring the OPERABLE DG from being paralled with the offsite network. However, this test must be performed every six months. This six month period provides adequate time for scheduling the SR without impacting DG OPERABILITY. ' Refer to the corresponding Bases for LC0 3.8.1 for a discussion of each SR. Also, refer to the Note before SR 3.8.11 and the Bases for a discussion of the Frequency requirements. I REFERENCES 1. -f4ft3(9tt%b Appendix 19.8A 4 - Amew ouien moww.16A Tech Spec Sms page g 1g 4 , l

1 System 80+ onion controlDocument l i l i Diesel Fuel Oil, Lube Oil, and Starting Air

 !                                                                                                                                                                         B 3.8.3
 !     BASES i                                                                                                                                                                                         -
 ;     SURVEILLANCE                SR 3.8.3.6                                                                                                                                              '

j REQUIREMENTS . j (continued) The draining of the fuel oil in the storage tanks, removal l 4 of accumulated sediment, and tank cleaning is required at i

 !                                 ten-year intervals by Regulatory Guide 1.137 (Ref. 2). This                                                                                                ;

also requires the performance of the ASME Code Section XI examinations of the tanks. To preclude the introduction of

surfactants in the fuel oil system, the cleaning should be _

accomplished using sodium hypochlorite solutions or their j equivalent rather than soap or detergents. This SR is for

preventive maintenance. The presence of sediment does not j necessarily represent a failure of this SR, provided that l accumulated sediment is removed during performance of the j Surveillance.

I REFERENCES 1. W Chapter 9. 1 2. Regulatory Guide 1.137, " Fuel Oil Systems for Standby Diesel Generators," October 1979.

3. ANSI N195-1976, " Fuel Oil Systems for Standby Diesel j -

Generators," Appendix B. i 4. ASTM Standards: D4057; D975; D4176; D1552; D2622; j D2276, Method A. 4 i - 1 i i i l i AnnrevedDeely, A000erial 16A Tech Spec Benee pay, g 3 g.57 i

Syotem 80+ Design ControlDocument l DC Sources - Operating i B 3.8.4 l BASES BACKGROUND Each battery charger has ample power-output capacity for the ' (continued) steady-state operation of connected loads required during normal operation while at the same time maintaining its battery bank fully charged. Each battery charger has sufficient capacity to restore the battery bank from the design minimum charge to 95% of its fully charged state in (12] hours while supplying normal steady-state loads .(Ref. 4). .. 1 APPLICABLE The initial conditions of design basis transient and

,    SAFETY ANALYSES          accident analyses in 5555145i!Oh, Chapters 6, Engineering             a 1

Safety Features, and 15, Accident Analyses, assume i Engineered Safety Features (ESF) systems are OPERABLE. The bDg L ) DC power subsystem provides normal and emergency DC power for the diesel generators, emergency auxiliaries, and for control and switching during all MODES of operation. The OPERABILITY of the DC power sources is consistent with the initial assumptions of the accident analyses which are basec

       ,                      upon maintaining the required DC power sources and associated distribution systems OPERABLE during accident i

conditions in the event of (1) an assumed loss of offsite AC ] power or all onsite AC power; and (2) a worst case single i failure. i The DC power sources satisfy Criterion 3 of the NRC Policy Statement. I 1 LC0 The Division 1 and 2 DC electrical power subsystems and corresponding control equipment and cabling are required to be OPERABLE to ensure availability of the required power to 1 ' j shutdown the reactor and maintain it in a safe condition 1 after an anticipated operational occurrence or a postulated i design basis accident. Loss of any one of the DC power subsystems does not prevent the minimum safety function from

 ,                            being performed. Each DC electrical power subsystem is considered OPERABLE if the 125 volt battery and associated battery charger satisfy the applicable Surveillance Requirements.

(continued) Apprend Design Meterial 16A Tech Spec Bases page g 2.g.go

Syntem 80+ oestan control Document DC Sources - Operating B 3.8.4 BASES REFERENCES 2. IEEE 308-1974, "IEEE Standard Criteria for Class 1E ' ' (continued) Power Systems for Nuclear Power Generating Stations."

3. 10 CFR 50 Appendix A, General Design Criteria 17,

                                                                            " Electric Power Systems."

4. M Chapter 8.

5. Regulatory Guide 1.93, " Availability of Electric Power Sources," December 1974.

6. IEEE 450-1980, "IEEE Recommended Practice for Maintenance Testing and Replacement of Large Lead Storage Batteries for Generating Stations and Subsystems."

7. Regulatory Guide 1.32, " Criteria for Safety-Related -

i Electric Power Systems for Nuclear Power Plants," February 1977.

8. IEEE 485-1983, " Recommended Practices for Sizing Large Lead Storage Batteries for Generating Stations and 6! _ C) (L, Substations." June, 1983.11

9. -d"4 Chapter 6. "

10. k Chapter 15. 4 l

I l l l l l l l l Apprend Deskn Meteriel 16A Tech Spec Bases pay, g 3,g.67

Sy tem 80+ Dessen controlDocument l DC Sources - Shutdown ' B 3.8.5 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.5 DC Sources - Shutdown - , BASES BACKGROUND A description of the DC Power Sources is provided in the Bases for LCO 3.8.4, "DC Sources Operating." APPLICABLE The initial conditions of Design Basis Accident (DBA) and

SAFETY ANALYSES transient analyses in MeHleWNWWE Chapter 6 and ' --

Chapter 154 assume that Engineered Safety Feature (ESF) systems are OPERABLE. The DC electrical power system provides normal and emergency DC electrical power for the DGs, emergency auxiliaries, and control and switching during all MODES of operation. ' The OPERABILITY of the DC subsystems is consistent with the ~ initial assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY. i ' The OPERABILITY of the minimum specified DC Power Sources during MODES 5 and 6 ensures that: 1) The plant can be maintained in the shutdown or refueling condition for extended time periods; 2) Sufficient instrumentation and control capability is available for monitoring and ' maintaining the unit status; and 3) Adequate DC electrical power is provided to mitigate events postulated during 4 shutdown, such as an inadvertent draindown of the vessel or j a fuel handling accident. 4 The DC sources satisfy Criterion 3 of the NRC Policy , Statement. LCO In MODES 5 and 6, the DC electrical power subsystems supporting the DC power distribution subsystem s of LCO 3.8.10, " Distribution Systems - Shutdown," are(re) quired to

.!                            be OPERABLE. This ensures the availability of sufficient power to recover from postulated events in MODES S and 6.

(continued) Approved Design Meterial 16A Tech Spec Bases pageg3.868

a j l Syntem 80+ Deslan control occument : " 1 DC Sources - Shutdown l j B 3.8.5 i 9 BASES (continued)

REFERENCES 1. % Chapter 6.

i ) 2. % Chapter 15.' -

                                                                            <     ~

i 3

                                                                                         ~~
 ?

I 4D 4 i 4 1 l b i WDeskrs Metodet.16A Tech Spec Resee Page B 3,8 71

                                                                  -   _.    ,.~ .

    . -             - = .     . _ - . _ -   .-     - -           _  --               -           .    .-  --

i System 80+ Desinn ControlDocument

Battery Cell Parameters j B 3.8.6 j BASES 4 BACKGROUND provide for correction of the measured values in accordance

1 (continued) with manufacturer's recommendations when the values reflect transient conditions as opposed to battery capacity.

Category C defines allowable values of electrolyte level,

,                          float voltage, and specific gravity of each connected cell.                            l i                           These values represent degraded battery conditions.

i However, operation is permitted when Category C limits are __ met since sufficient capacity exists to perform the intended function. These values are discussed in more detail in the l ACTIONS section of this Bases. 1 I , 2 i APPLICABLE The initial conditions of design basis transient and j i SAFETY ANALYSES accident analyses in tO Emet $. Chapters 6, Engineering - l Safety Features, and 15, Accident Analyses, assume all t Engineered Safety Features (ESF) systems are OPERABLE. The Q _ D C_ DC electrical power systems provide normal and emergency DC i power for the DGs, emergency auxiliaries, and control and ]' switching during all MODES of operation. The OPERABILITY of the DC subsystem is consistent with the initial assumptions of the accident analyses and is based upon maintaining one ! division of DC power sources and associated distribution ! systems OPERABLE during accident conditions in the event of (1) an assumed loss of all offsite AC power or all onsite AC l power; and (2) a worst case single failure. l Battery cell parameters satisfy Criterion 3 of the NRC

Policy Statement.

I 4 LCO Battery cell parameters must remain within acceptable limits to ensure availability of the required DC power to shut down the reactor and maintain it in a safe condition after an anticipated operational occurrence or a postulated DBA. .' Electrolyte limits are conservatively established, allowing 1 continued DC electrical system function even with Category A i 1 and B limits not met. i i i l i ) l (continued) ? 4pwond Du> Atesodel- 16A Tech Spec Buu Page 3 3.8 73 ]-

i Syntem 80+ oestan controlDocument Battery Cell Parameters B 3.8.6 BASES (continued) REFERENCES 1. IEEE 450-1980, "IEEE Recommended Practice for ' Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations."

2. IEEE 308-1978, "IEEE Standard Criteria for Class IE Power Systems for Nuclear Power Generating Stations

3. M Chapter 6.

4. Chapter 15. 4 I

P l Approved Deelers Material 16A Tech Spec Bases pageg 3.g.go

Sv tem 80 + Declan ContmlDocument Inverters - Operating B 3.8.7 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.7 Invertersr-Operating , BASES BACKGROUND The inverters are the preferred source of power for the AC vital buses becaus? of the stability and reliability they

,                                                    achieve in being powered from the 120 VDC battery source.

The function of the inverter is to convert DC electrical - 1 power to AC electrical power, thus providing an - - l uninterruptible power source for the instrumentation and

,                                                    controls for the Reactor Protective System (RPS) and the Engineered Safety Feature Actuation System (ESFAS).

Specific details on inverters and their operating characteristics are found in @ Chapter 8 4 -- (Ref. 1). (L D C-APPLICABLE The initial conditions of Design Basis Accident (D8A) and j SAFETY ANALYSES transient analyses inn Chapter 6 (Ref. 2) and i

Chapter 15 (Ref. 3)g assume Engineered Safety Feature systems are OPERABLE. The DC to AC inverters are designed 4 to provide the required capacity, capability, redundancy, j and reliability to ensure the availability of necessary power to the RPS and ESFAS instrumentation and controls so that the fuel, Reactor Coolant System, and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for Section 3.2, Power l i Distribution Limits; Section 3.4, Reactor Coolant System (RCS); and Section 3.6, Containment Systems. The OPERABILITY of the inverters is consistent with the initial assumptions of the accident analyses and is based on meeting the design basis of the unit. This includes maintaining required AC vital buses OPERABLE during accident

conditions in the event of

l

a. An assumed loss of all offsite AC electrical power or all onsite AC electrical power; and

b. A worst case single failure.

Inverters are a part of the distribution system and, as j such, satisfy Criterion 3 of the NRC Policy Statement. l l l' (continued) 4prevenf Denfpr Moseriel.16A Tech Spec Resee pope B 3.8-81

I System 80+ Destan control Document Inverters - Operating 8 3.8.7 BASES B.1 and B.2 ACTIONS (Continued)

,                              If the inoperable devices or components cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LC0 does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours and to MODE 5 within 36 hours. The allowed Completion Times are reasonable,                  __

based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner

;                             and without challenging unit systems.

i SURVEILLANCE SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter. The verification of proper voltage and frequency output ensures that the required power is readily available for the instrumentation of the RPS and ESFAS connected to the AC vital buses. The 7 day Frequency takes into account the redundant capability of the inverters and other indications available in the control room that alert the operator to inverter malfunctions. l REFERENCES 1. M Chapter 8. 4 ~ l 2. 'EOQR:3K. Chapter 6. 4 _

3. % Chapter 15. c I

l J Approved Desierr Materiel 16A Tech Spec Bases pay, g 3 g.g4 W

Sv;-tem 80+

Deslan controlDocument l 1 Inverters - Shutdown j B 3.8.8 ; } B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.8 Inverters-Shutdown - j , j . . j BASES i 4 BACKGROUND A description of the inverters is provided in the Bases for j LC0 3.8.7, " Inverters-Operating." q APPLICABLE The initial conditions of Desi n Basis Accident (DBA) and j

SAFETY ANALYSES transient analyses in Chapter 6 (Ref. 1) and 4 - Chapter 15 (Ref. 2)(assume Engineered Safety Feature ' j systems are OPERABLE.' The DC to MC inverters are designed ! to provide the required capacity, capability, redundancy, j and reliability to ensure the availability of necessary j power to the Reactor Protective System and Engineered Safety Features Actuation System instrumentation and controls so that the fuel, Reactor Coolant System, and containment ~ l , design limits are not exceeded. 4 l

             '                               The OPERABILITY of the inverters is consistent with the j                                             initial assumptions of the accident analyses and the

requirements for the supported systems' OPERABILITY.

i j The OPERABILITY of the minimum inverters to each AC vital bus during MODES 5 and 6 ensures that:

a. The unit can be maintained in the shutdown or j refueling condition for extended periods; ,

i b. Sufficient instrumentation and control capability is { available for monitoring and maintaining the unit f status; and i

c. Adequate power is available to mitigate events postulated during shutdown, such as an inadvertent j

draindown of the vessel or a fuel handling accident. l The inverters were previously identified as part of the

distribution system and, as such, satisfy Criterion 3 of the j j NRC Policy Statement. '

i l i i 1 i

i (continued) j herewaro elvo ne enew- 1sA rus so csua rene a 2.sts 4

System 80+ oesion controlDocument Inverters - Shutdown B 3.8.8 l BASES (continued) SURVEILLANCE SR 3.8.8.1

REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter. The verification of proper voltage and' frequency output ensures that the required power is readily available for the-instrumentation connected to the AC vital buses. The 7 day ..

Frequency takes into account the redundant capability of the inverters and other indications available in the control room that alert the operator to inverter malfunctions. REFERENCES 1. K Chapter 6. G

2. Chapter 15. g

'1 l t 1 i 1 1 ! 1 l

i a Anoroved Destgre Materiel 16A Tech Spec Bom p,g, g 3,g.gg \ -l

i j System 80+ Deslan CrntrolDocument I Distribution Systems - Operating ! . B 3.8.9 ' i ! BASES { . . lBACKGROUNO The list of all required distribution buses is located in *

f (continued) Table B 3.8.9-1. APPLICABLE ,

(. D L l APPLICABLE The initial conditions of design basis transient and SAFETY ANALVSES accident analyses in **

Chapters 6, Engineering <

Safety Features, and 15, Acciilent Analyses, assume Engineered Safety Features (ESF) systems are OPERABLE. The

i. AC, DC, and AC Vital Electrical Power Distribution Systems
are designed to provide sufficient capacity, capability, redundancy and reliability to ensure the availability of necessary power to ESF systems so that the fuel, Reactor 3

Coolant System, and containment design limits are not exceeded. These limits are discussed in more detail in the Bases for LCO Sections 3.2. (Power Distribution Limits), 3.4 (Reactor Coolant System), and 3.6 (Containment Systems). , I The OPERABILITY of the Electrical Power Distribution Systems is consistent with the initial assumptions of the accident

;               .      analyses and are based upon maintaining at least one of the J                       onsite AC, DC, and Vital AC power sources and associated i                     distribution systems OPERABLE during accident conditions in i                       the event of (1) an assumed loss of all offsite power or all

! onsite AC power, and (2) a worst case single failure. , The distribution systems satisfy Criterion 3 of the NRC l Policy Statement. 1 LCO The Power Distribution System Divisions listed in Table B 3.8.9-1 ensure the availability of AC, DC, and Vital AC Electrical power for the systems required to shutdown the reactor and maintain it in a safe condition after an i anticipated operational occurrence (A00) or a postulated I design basis accident. Two divisions of the AC, DC, and AC ! Vital Electrical Power Distribution Systems are required to

be OPERABLE.

I Maintaining two divisions of AC, DC, and AC Vital Electrical Power Distribution Systems OPERABLE ensures that the redundancy incorporated into the design of ESF is not l defeated. Either division of the distribution system is t capable of providing the necessary electrical power to its (continued) l w'omar, neeeuw. ssa run seu seu n eo a 2.s-so

Syntem 80+ oesten control Document

,                                                                           Distribution Systems - Operating B 3.8.9 i

i BASES (continued) l

;                                                                                                                                      1 i                                                                                                                                  .

i REFERENCES 1. Regulatory Guide 1.93, " Availability of Electric Power

l Sources," December 1974. 1 l

2. Chapter 6. m

3. Chapter 15.

i i l 1 l l ) i k i ?, i 4 I 1 i 1 4 Anerend Du> MetwW.16A Tech Spec Buu 9 g 3.g.gy I

System 80+ oeslan controlDocument l l Distribution Systems - Shutdown l , B 3.8.10 l

' B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.10 Distribution Systems - Shutdown                                                                    ,. ,

) . . BASES 1

' BACKGROUND            A description of the AC, DC, and AC Vital Power Distribution Systems is provided in the Bases for 'LCO 3.8.9,
,                       " Distribution Systems - Operating."                                                        l

) ! d DC I i APPLICABLE The initial conditions of Design Basis Accident and ! ! SAFETY ANALYSES transient analyses in W Chapter 6 (Ref.1) and 4 l l Chapter 15 (Ref. 2)g assume Engineered Safety Feature (ESF) 1 systems are OPERABLE. The AC, DC, and AC vital bus

!                       electrical power distribution systems are designed to 4

provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to

                                                                                                                    )
)                       ESF systems so that the fuel, Reactor Coolant System, and                                   i containment design limits are not exceeded.

1 4 , The OPERABILITY of the AC, DC, and AC vital bus electrical power distribution system is consistent with the initial

~ assumptions of the accident analyses and the requirements for the supported systems' OPERABILITY. i The OPERABILITY of the minimum AC, DC, and AC vital bus i electrical power distribution subsystems during MODES 5 l and 6 ensures that: ) i a. The unit can be maintained in the shutdown or l refueling condition for extended periods; l

b. Sufficient instrumentation and control capability is !

available for monitoring and maintaining the unit I status; and I 1

c. Adequate power is provided to mitigate events postulated during shutdown, such as an inadvertent j draindown of the vessel or a fuel handling accident.

The AC and DC electrical power distribution systems satisfy Criterion 3 of the NRC Policy Statement. l (continued)

%w w useaw. tea rech sm sua                                                     reo, a 2.s se

evstem 80+ Design ControlDocument i Distribution Systems - Shutdown B 3.8.10 BASES (cor tinued) t - . h0RVEILLANCE SR 3.8.10.1 ' ' '

fQUIREMENTS This Surveillance verifies that the AC, DC, and AC vital bus electrical power distribution system is functioning properly, with all the buses energized. The_ verification of 4 proper voltage availability on the buses ensures that the . I required power is readily available for motive as well as !

control functions for critical system loads connected to . .. l these buses. The 7 day Frequency takes into account the redundant capability of the electrical power distribution subsystems, and other indications available in the control room that alert the operator to subsystem malfunctions. hFERENCES 1. % Chapter 6. I

2. Chapter 15.

m nwendDee&n Meterkt 16A Tech bec Basee Page B 3.8 102

i i

System 80+ w controlDocument 1
Boron Concentration i

i B 3.9.1 l l BASES t 1 l ACTIONS A_J (continued) *

d Once boration is initiated, it must be continued until the

boron concentration is restored. The completion time depends on the amount of boron which must be injected to 1 l reach the required concentration.

SURVEILLANCE SR 3.9.1.1 i REQUIREMENTS ' This SR ensures the reactor coolant boron concentration in the RCS, refueling canal and refueling cavity is within the COLR limits. The boron concentration in the coolant is determined periodically by chemical analysis. ~ j Because the likelihood of a significant reduction in the

boron concentration during MODE 6 operations is remote, a' -: minimum frequency of once every 72 hours is a reasonable 4 interval to verify boron concentration. The surveillance ,

j interval is based on extensive operating experience and 1 ensures that the boron concentration is checked at adequate { intervals. REFERENCES 1. 10 CFR 50, Appendix A, Section VI, Criterion 26, i

                            " Reactivity Control System Redundancy and Capability."

I. 2. NS-51.2, ANSI /ANS-57.2-1983, Section 6.4.2.2.3, i American Nuclear Society, American National Standard, i " Design Requirements for Light Water Reactor Spent ! Fuel Storage Facilities at Nuclear Power Plants," i 1983. f a 3. M % Chapter 15. 4 j 4. 52 FR 3788, NRC Interim Policy Statement, on Technical l Specification Improvements for Nuclear Power Reactors, j February 6,1987. 1 (continued) I, i iww u eaw rsa rus seu nau r.o. s 2.u i

l System 80+ Deslan Control Document 1 Boron Concentration B 3.9.1 BASES REFERENCES 5. NRC Bulletin No. 89-03, " Potential Loss of Required - (continued) Shutdown Margin During Refueling Operations," November 9,/ek 21, 1989. .,

                                                                                               'eessee- 00,

6. Section 19.8A, Shutdown Risk Evaluation 2 h le k w oc d 9

Seper$ I I Anwend Design Materiel 16A Tech Sm Bases p ,,,y 3,y.5 l

i ! [ . System 80+ __ Design C=ntrol Document ' NLclear Instrumentation B 3.9.2 BASES (continued) i - SURVEILLANCE SR 3.9.2.1 * ' i REQUIREMENTS j SR 3.9.2.1 is the performance of a CHANNEL CHECK, which is 4 the comparison between channels of the indicated parameter values for each of the functions. It is based on the

assumption that the two indication channels should be consistent with core conditions. Changes in fuel loading i

and core geometry can result in significant differences between source range channels but each channel should be ,, i consistent with its local conditions. The Frequency of 12 hours is based on the importance of the SRMs. The Frequency j is consistent with LCO 3.3.5, " Engineered Safety Features l Actuation System (ESFAS) Instrumentation" and has been j proven acceptable through operating experience. 1 i . SR 3.9.2.2 SR 3.9.2.2 is the performance of a CHANNEL CALIBRATION every 18 months. This SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. 1 The CHANNEL CALIBRATION for the SRMs consists of obtaining j the detector plateau or } evaluating those curves, preamp discriminator and comparing the curvescurves, to the manufacturer's data. The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage. Operating experience has shown that these components usually pass the Surveillance when l performed on the 18 month Frequency. 4 I jREFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. h Chapter 15. _

4 i ) { i a j i l i 1 , AnorovedDesign Meterial- 16A Tech Spec Rosen Page B 3.9-8

__ .- _._ . - . ~ . . _ . _ .._ .- _ . _. - _ _ _ _ . i System 80+ Design ControlDocument Containment Penetrations B 3.9.3 BASES j l 5 $ j BACKGROUND The other containment penetrations that provide direct i (continued)- access from containment atmosphere to outside atmosphere j must be isolated on at least one side. Isolation may be i achieved by an OPERABLE automatic isolation valve, or by a

manual isolation valve, blind flange, or equivalent. 1 1 Equivalent isolation methods must be approved and may !

j include use of a material that can provide a temporary, atmospheric pressure ventilation barrier for the other j containment penetrations during fuel movements (Ref. 2). i i , } APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel 1 SAFETY ANALYSIS assemblies within containment, the most severe radiological l consequences result from a fuel handling accident. The fuel 4 handling accident is a Condition IV postulated event which ! involves damage to irradiated fuel (Ref. 3). Fuel handling j accidents, analyzed in %E654RdhSection 15.7.3.4, include d dropping a single fuel assembly and handling tool or a heavy 0, D] ] object onto other irradiated fuel assemblies (Ref. 4). The

requirements of this LCO and LC0 3.9.6, " Refueling Water j Level", and the minimum decay time of [72] hours prior to 1 CORE ALTERATIONS ensure that the release of fission product t

radioactivity subsequent to a fuel handling accident does j not result in doses in excess the guideline values specified i in 10 CFR 100 and Standard Review Plan Section 15.7.4, Rev. j 4 1 (Ref. 3). l Containment enetrations satisfy Criterion 3 of the NRC i Policy Statement. 4 lLCO This LC0 limits the consequences of a fuel handling accident ; j in containment by limiting the potential escape paths for 4 fission product radioactivity released within containment. 1 i The LCO requires any penetration providing direct access i L from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge and i exhaust penetrations. l For the OPERABLE containment purge penetrations, this LCO i j ensures that these penetrations are isolable by the

l Containment Isolation Actuation Signal. The OPERABILITY l i

t requirements for this LCO ensure that the automatic purge I i { (continued) l l 1 i AapresedDenkre nieenriel 16A Tech Spec senee page g 3.g.g y 4 )'

! Syntem 80+ Desian Control Document l Containment Penetrations B 3.9.3 i BASES ]

04 W /d* t LC0 and exhaust valve closure times specified in * '

(continued) can be achieved and therefore meet the assumptions used in g c0 the safety analysis to ensure releases through .the valves are terminated, such that the radiological doses are within j the acceptance limit. 1 1 i l

APPLICABILITY The containment penetration requirements are applicable . during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment since this is when there is a i potential for a fuel handling accident. In MODES 1, 2, 3 and 4, Containment Penetration requirements are addressed by i LCO 3.6.1, " Containment". In MODES 5 and 6 when CORE l ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment -

penetration status.

I

ACTIONS A.1 and A.2 With the containment equipment hatch, air locks, or any containrant penetration providing direct access from the 1 containment atmosphere to the outside atmosphere not in the .

required status, including the Containment Purge and Exhaust I system not capable of automatic actuation when the purge and l

!                  exhaust valves are open, the unit must be placed in a 2

condition where the isolation function is not needed. This is accomplished by immediately suspending CORE ALTERATIONS

and movement of irradiated fuel assemblies within containment. Performance of Required Actions A.1 or A.2 i shall not preclude completion of actions to establish a safe condition.

l 1 SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This SR verifies that each of the containment penetrations

'                  required to be in its closed position is in that position or is capable of being closed by an OPERABLE automatic Containment Isolation Actuation Signal.      As such, this (continued) w o e u. w. m r- s,,-                                                  ,,,,. e .,. ,. n

i , System 80+ Deslan ControlDocument i ! Containment Penetrations j 8 3.9.3 4 l 8ASES - j SURVEILLANCE SR 3.9.3.1 (continued) '

*i

, REQUIREMENTS i Surveillance ensures that a postulated fuel handling 4 accident which involves a release of fission product radioactivity within the containment will not result in a j release of fission product radioactivity to the environment. ; ! The SR is performed every 7 days during CORE ALTERATIONS or ~ movement of irradiated fuel assemblies within containment. l i The Surveillance interval is based on the importance of ! these penetrations to restrict the release of fission product radioactivity to the environment and has been shown to be acceptable through operating experience. i SR 3.9.3.2 ^ This SR demonstrates each containment purge and exhaust ~ valve actuates to its isolation position on an actual or simulated actuation signal. The [18 month] Frequency maintains consistency with similar ESFAS testing requirements and has been shown to be acceptable through operating experience. REFERENCES 1. 10 CFR 20, Standards For Protection Against Radiation. i 2. "Use of Silicone Sealant to Maintain Containment Integrity - ITS", GPU Nuclear Safety Evaluation SE-l 0002000-001, Rev. O, May 20, 1988.

3. NUREG-0800, Standard Review Plan Section 15.7.4, Radiological Consequences of Fuel Handling Accidents Rev. 1, July 1981.

4. Ypi$Af Chapter 15.

I i e 4 proved Design Mowiel- 16A Tech Spec genee pope a 3.s.12

System 80+ &b contml Document SCS and Coolant Circulation - High Water Level , B 3.9.4 4 BASES ACTIONS $ A_J -l (continued)

If SCS division requirements are not met, immediate actions l shall be taken and continued to satisfy the SCS division requirements. With the unit in MODE 6 and the refueling I cavity water level it 23 feet above the top of the reactor '

i vessel flange, the completion time of immediate ensures that

prompt action is taken to meet the necessary SCS division cooling requirements. _

L.4 If SCS division requirements are not met, all containment " penetrations to the outside atmosphere must be closed to prevent fission products, if released by a loss of decay heat event, from escaping the containment. The 4 hour Completion Time provides ample opportunity to fix SCS -

problems without incurring the additional action of violating the containment atmosphere.

SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance verifies that the SCS division is operating and circulating reactor coolant. Tie flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. The Frequency of 12 hours is sufficient considering the flow, temperature, pump control, and alarm indications available to the operator to monitor the SCS in the control room. This Frequency ensures that SCS division operation and flow is checked at adequate i intervals. i , i

)

REFERENCES 1. Chapter 5. <

2. Chapter 15. -

t l l l

l (continued)

Anwovee Day Moserw.1sA Tech spec senee i rope s .1.s.17 1

System 80+ Deslan contror Document - SCS and Coolant Circulation - High Water Level B 3.9.4 BASES

                                                                                                                    .  ,i REFERENCES                3.      "NRC Staff Review of Nuclear Steam Supply Vendor                                -  '

(continued) Owners Groups' Application of the Commission's Interim Policy Statement criteria to Standard Technical Specifications," transmitted by Thomas E. Murley (NRC) letter to Joseph K. Gasper (CE0G) dated May 9, 1988.

4. 52 FR 3788, NRC Interim Policy Statement, on Technical Specification Improvements for Nuclear Power React February 6, 1987. Dcle k -

u

5. Section 19.8A, Shutdown Risk Evaluation T E'5" #3 1

pe Ie k Luo<xl Se p r ?;,:eved Des &rs Meterial 16A Tech Spec Bases page a 3,s.pg

System 80+ Design ControlDocument l l SCS and Coolant Circulation - Low Water Level l B 3.9.5 -BASES SURVEILLANCE SR 3.9.5.2 ~

REQUIREMENTS -

(continued) Verification is performed by ensuring correct breaker alignment and indicated power available to the required pumps. The Frequency of seven days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. SR 3.9.5.3 - Verification of the correct breaker alignment and indicated power available to the operable CS pump ensures that the CS pump will be able to remove heat from the RCS in the event of a power failure to the operating SCS division. The Frequency of [24 hours) is based on operating experience. REFERENCES 1. M Chapter 5. ~

2. M Chapter 15. <

3. "NRC Staff Review of Nuclear Steam Supply Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications," transmitted by Thomas E. Hurley (NRC) letter to Joseph K. Gasper (CE0G) dated May 9, 1988.

4. 52 FR 3788, NRC Interim Policy Statement, on Technical Specification Improvements for Nuclear Power Reactors, February 6, 1987.

5. M Section 19.8A, Shutdown Risk Evaluation De lo {c.

u Q 'e c ss AK ut , D e l e N *** 4 ka t** b l pprovedDesign Material- 16A Tech Spec Bases ! Pope B 3.9 23

System 80+ Design ControlDocument Refueling Water Level B 3.9.6 i I BASES SURVEILLANCE SR ?.9.6.1 (continued) REQUIREMENTS the consequences of a postulated fuel handling accident inside containment which results in damaged fuel rods (Ref. 2). The 24 hour Frequency ensures that the water is at the required level and is considered adequate due to the large ~

volume of water and the normal procedural controls of valve j positions, significant unplanned level changes are unlikely.

i REFERENCES 1. USNRC Regulatory Guide 1.25, Assumptions Used for . Evaluating the Potential Radiological Consequences of 2 a Fuel Handling and Storage Facility for Boiling and i Pressurized Water Reactors, March, 1982. i '

2. M Chapter 15.

i i

3. 52 FR 3788, " Proposed Policy Statement on Technical Specifications Improvements for Nuclear Power Plants",

February 6, 1987.

4. 10 CFR 100.11, " Determination of Exclusion Area, Low i

Population Zone and Population Center Distance." i l' I

5. NUREG-0800, " Standard Review Plan", Section 15.7.4 ;

} Radiological Consequences of Fuel Handling Accidents, U.S. Nuclear Regulatory Commission. ] 4 I f i Approved Deelen Metodel- 16A Tech Spec Bases y,y, y 3,y.gg

1 CESSAR !!nlam sk j i 76 % vwGG I 4 1 { operation. Maintenance that will be performed more frequently than i refueling outages must be planned so as to not disrupt operation or j be likely to cause reactor scram, engineered safety feature (ESF) i actuation, or abnormal transients. Maintenance, planned for performance during refueling outages, must be conducted in such a }i way that it will have little or no impact on plant safety, on i outage length, or on other maintenance work. i 1 As plant experience data accumulates, the failure rates and human I

l error rates in the operations reliability assurance process should -

)                be updated.       These failure rates were used by the designer in the PRA.

1 The COL applicant should provide a complete operations reliability [Fassurance process description to be reviewed by the NRC (see M M Section17.3.13)][COLItem17-3] { l 17.3.10 OPERATIONS RELIABILITY ASSURANCE PROCESS operations reliability assurance process that is expected to be j prepared and implemented by the COL applicant [ COL Its 7-3), j should make use of the information provided by the designe This

 !               information will help the owner / operator determine activiti that

] should be includedin the operations _ reliability assurance process. [_ ^ ! l , pamplesofelementsthatmightbeincluded]areasfollows: Reliability Performance Monitoring -- Measurement of the j ib ( performance of equipment to determine that it is accomplish'ing pc \ its goals and/or that it will continue te operate with/ low , ,gs 1 .kt t [ should preferably be predictive in nature to prevent probability of failure and high availability. Monityring loss of

                                                                                                               $ ,d g !

lp critical functions. .Y I Reliability can compareMethodology plant data to -- Methods the SSC databy inwhich the plant /operytorhh the PRA. ) j Problem Prioritization -- Identification, for each of the j risk- significant SSCs, of the importance of that item as ! contributor to its system unavailability and assignment of priorities to problems that are detected with such equipment. {

Root Cause Analysis -- Determination, for problems that occur l

' regarding reliability of risk-significant SSCs, of the root g i causes, those causes which, after correction, will not recur / to again degrade the reliability of equipment. / Dud i Corrective Action Determination -- Identification of corrective actions needed to restore equipment to its required j functional capability and reliability, based on the results of q problem identification and root cause analysis. l Corrective Action Implementation -- Carrying out identified

corrective action on risk-significant equipment to restore

! equipment to its intended function in such a way that plant j safety is not compromised during work. 1

Amendment W
17.3-10 June 17, 1994 l

1 ._ _ -_. _.

j i CESSARSH hn.s i Corrective notion verification -- Post-corrective action tasks to be followed after maintenance on risk-significant equipment to assure that such equipment will perform its intended l functions. l l Plant Aging. -- Some of the risk-significant equipment- is expected to undergo age related degradation that will require equipment replacement or refurbishment.

Feedback to Designer --

The plant owner / operator should i periodically compara performance of risk-significant equipment

to that specified in the PRA and D-RAP, and, at its f discretion, may send SSC performance data to plant or equipment designers in those cases that consistently show performance below that specified. The plant owner / operator !

j should consider participation in the CEOG. ' l Programmatic Interfaces -- Reliability assurance interfaces ; y related to the work of the several organizations and personnel ' l groups working on risk-significant SSCs.

unintenance Rule Integration --

The plant owner / operator should consider the integration or interface of operations reliability assurance process and the requirements of 10 CFR i 50.65 which require the operator to develop a maintenance i program for risk significant SSCs or SSCs that could produce l j trips or transients, pg [Theplantowner'soperationsreliabilityassuranceprocessshould address interfaces : the with construction, startup testing, pl cperations, maintenance, engineering, safety, licensing, qualiti y y < essuranceandprocurementofreplacementequipment.] pH , 17.3.11 r l D-R1P IMPLEMENTATION ) i An example of implementation of the D-RAP is given for the i i Component cooling Water System (CCWS). This system was selected as I on example because it was a support system and was found in the carlier l System 80 PRA to contain risk-significant components. Because of this finding, and through the D-RAP organization described in Section 17.3.5, the design was changed. The design ! cnd analytical results, as presented in this chapter, is presented i only as a D-RAP example and does not necessarily correspond to the j t current System 80+ design, l 17.3.11.1 CCWS Function i ' The component Cooling Water System (CCWS) is a closed loop system that provides cooling water flow to remove heat released from plant cystems, structures, and components. The CCWS functions to cool the safety-related and non-safety-related reactor auxiliary loads. j Heat transferred by these components to the CCWS is rejected by the Station Service Water System (SSWS) via the CCWS heat exchangers.

b l Amendment W i 1 17.3-11 June 17, 1994

I System 80+ Design ControlDocument Table 18.5.2-3 General System 80+ Component Data Component Desi;tn & Operating Data 4;ESSAR-DC A -CE55aR-F ' ~~ l Component f Mystem 80+P P NSystem 80V4 l l Reactor Vessel l l Total core heat output (MWt) 3,914 3,800 Design pressure (psia) 2,500 2,500 Primary system pressure (psia) 2,250 2,250 RCS inlet temperature ('F) 4M-& 6% 66t$ 'i& S RCS outlet temperature (*F) 615 621 Design minimum RCS flow rate (gpm) 444,650 445,600 Steam Generator Number of units 2 2 Primary Side Design pressure (psia) 2,500 2,500

                   . Design temperature (*F)                                   650                       650 Operating pressure (psia)                                2,250                     2,250 Secordary Side Design pressure (psia)                                   1,200                     1.270 Design temperature (*F)                                   570                       575 Full Load Steam Pressure (ps a)                          1,000                     1,070             1 Full Load Steam Temperature (*F)                          545                       553 Zero Load Steam Pressure (psia)                          1,100                     1,170 Total Steam Flow per gen. (Ib/h)                       8.82x106                  8.59x106 Full load steam quality (%)                              99.75                     99.75 Pressurizer Internal free volume (fd)                                2,400                     1,800 Design pressure (psia)                                   2,500                     2,500 Design temperature (*F)                                   700                       700 Operating pressure (psia)                                2,250                     2,250             l Operating temperature (*F)                                653                       653 Pressurizer j                      Vessel height (ft) Component                                54                        42 l

1 Approved Design Material- Human Factors Engineening Page 18.516 i

h E RCP1A RCP1B

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