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Corrected Nonproprietary Facility Conceptual Design Description for Technical Support Ctr & Emergency Operations Facilities.
	ML17331A802
Person / Time
Site:	Cook
Issue date:	09/14/1981
From:	; INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	;
Shared Package
ML17331A800	List: ML17331A799; ML17331A801; ML17331A802;
References
	NUDOCS 8109230495
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{{#Wiki_filter:INDIANA & MICHIGAN EUKTRIC COMPANY DONALD C. COOK NUCHRR PLANT FACILITY COIKEPTUAL DESIGN DESCRIPTION FOR THE TECHNICAL SUPPORT CENTER AND THE EOF. ATTACHMWZ TO AEP:NRC: 0531C This docum nt contains information proprietary to Anglican Electric Power Service Corporation; for it is submitted in confidence and is to be for it is furnished. This docum nt used solely the purpose which and such information is not to be reproduced, transmitted, disclosed or used otherwise in whole or in part without authorization of Am rican Electric Power Service Corporation. 8109230495 810814 PDR ADOCK 05000315 PDR

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This document contains mater ial that is proprietary to the Westinghouse 0- Electric Corporation. The proprietary information has been marked by means of brackets. The basis for marking the material proprietary is identified by marginal notes referring to the standards in Section 8 of the'affidavit of R. A. Wiesemann of record "In the Hatter of Acceptance Criteria for Emergency Core Cooling Systems for Light Mater Cooled Nuclear Power Reactors (Oocket No. RH-50-1)" at transcr ipt pages 3706 through 3710 (February 24, 1972). Oue to the proprietary nature of the material contained in this report which was obtained at consider able Westinghouse expense and the release of which would seriously affect our competitive position, we request this information to be withheId from public disclosure in accordance with the Rules of Practice, 10 CFR 2.790, and that the information pre-sented therein be safeguarded in accordance with 10 CFR 2.903. We believe that withholding this information will not adversely affect the public interest.- This information is for your internal.use only and should not be released to persons or organizations outside the Oirectorate of Regula-tion and the ACRS without prior approval of Westinghouse Electric Corporation. Should it become necessary to release this information to. such persons as part of the review procedure, please contact Westing-house Electric .Corporation and they will make the necessary arrangements required to protect their proprietary interests.

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j TABLE OF CONTENTS I

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I Section Title Pacae Introduction AEP-1 System Functions AEP-1 1.1.1 Technical Support Center AEP-1 1.1.2 Safety Parareters Display System 1.1.3 Nuclear Data Link 1.1.4 +pass & Ingle Status Indication System 1.2 Report Basis

2. The Data Acquisition & Display System AEP-4 2.1 Computer System 2.2 Input System 2.3 Data Display System 2.3.1 Ohsite Technical Support Center AEP-5 2.3;2 Contxol Rocm AEP-5 2.3.3 Emergency Operating Facilities

3. Onsite Technical support Center AEP-9 3.1 Design Basis AEP-9 3.2 Input D termination AEP-10 3.3 (ZSC Operator Interface AEP-11 4~ Safety Paraneters Display System AEP-30 4.1 Purpose AEP-30 4.2 Input Detexmination AEP 30 4.3 Man-iMchine Interface AEP-33

5. Bypass & Inoperable Status Indication System AEP-47 5.1 Purpose AEP-47 5.2 Input Determination AEP-47 5.3 Pan-Machine Interface AEP-47

6. TSC Instxum ntation AEP-55

TABLE OF CONTENTS Section. Title Pacae

7. TSC Power Supply Systems AEP-56 7.1 Power to the TSC Computer AEP-56 7.1.1 'Ihe UPS System AEP-56 7.1.2 Consequence of Power Supply AEP-56 7.2 ~ to the Interxuption TSC Complex AEP-57

8. TSC and EOF Functions AEP-58 8.1 Task Functions Performed by Individuals AEP-58 in the TSC.

8.1.1 Radiation Monitoring AEP-58 8.1.2 Ebse Assessnant AEP-58 8.1.3 Gmnunications AEP-58 8.1.4 Technical Support AEP-59 8.1.5 Management Support AEP-59 8.2 Bnergency Functions Performed in the AEP-59 TSC/EOF for each Emergency Class. 8.2.1 &usual Event AEP-59 8.2.2 Alert AEP-60 8.2.3 Site and General Emergency AEP-61 8.3 Functions of Individuals Reporting AEP-62 to the EOF.

9. TSC Records and Data Availability AEP-63 9.1 Controlled Plant Specific Reference Material AEP-63 9.2 &controlled Information and Technical AEP-64 Reference Material.

9.3 Other Data, Records, and Information AEP-65

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1. INTB3DKZION

1. 1 SYSTEM FUNCTIONS:

The D.C. Cook Plant Technical Support Center Data System is being developed and designed using the guidelines of NUREG 0696 to provide the plant cperating and technical support personnel with the pertinent plant information to facilitate the errergency response to an accident. This System, which utilizes the Westinghouse P2500 TSC Computer Systems, can also be used during normal plant operation for other functions such as plant performance analysis, personnel training etc. This system consists of two similar computerized data acquisition, processing and display systems, one for each D.C. Cook Unit. The four najor functions prcvided by this computer system are: 1.1.1 TECZKICAL SUPPORT CENTER (TSC): The ~ter system will receive, store, process and display on color CRT terminals and/or on hard-copy terminals the real time data acquired from various plant systems. Pre-trip and post-trip data are also collected and can be processed and displayed by the computer. This system will facilitate the assesmeat of the plant's condition by plant operating and technical support personnel. The data displays of the Technical Support Center function will provide sufficient information to determine: AEP-1

~ ~ \ Plant steady state cperating conditions prior to the unit trip.' Transient conditions producing the initiating event and system behavior during the course of the accident. Present conditions of the plant . The TSC data display system may be used for: Reviewing the accident sequence. Determining apprcpriate mitigating actions. Evaluating the extent of any damage. Determining plant status during recovery operations. This function will be described in details- in Section 3. 1.1.2 PLANT SAFETY STATES DISPLAY (PSSD):

     ~s    PSSD  system was designed       in  accordance    with the guidelines for  the Safety Parameter        Display System      (SPDS)   of NUREG   0696.

This PSSD system, which displays the safety status of the plant in a format that can be easily recognized by the contxol room operators, will help the operators to detect any abnormal condition in a timely manner. Additional features of this PSSD system will help the operators and technical support personnel to obtain detailed information on the safety systems of the plant. Detailed descriptions of this system are provided in Section 4. 1.1.3 NKXZAR DATA LINK (NDL) The TSC computer system has a built-in off-site data transmission capability which can be used for interfacing with a future Nuclear Data Link (NDL) Sub-System. AEP-2

1.1.4 BYPASS & INOPERABLE STATUS INDICATION SYSTEM (BISI): The BISI system provides the operators and technical support personnel with a clear indication of the availability of the plant safety systems (ESP Systems). Detailed descriptions of this system are provided in Section 5. 1.2 REPORT BASIS: This report is based on the proprietary Westinghouse KM? Report 9725 Westinghouse Technical S~rt Complex" which was submitted to the NRC. Appropriate nedifications were made to reflect the specific design of D.C. Cook M.ts 1 and 2. AEP-3

2. THE IRTA AOQUISITION & DISPLAY SYSTEM

2. 1 THE CGMPUZER SYSTEM:

Figure 2.1 shows the ccmputer system hardware for each Cook Unit. Multiple 16-bit high speed minicomputer and merry devices are used to process plant data, generate displays and perform other man-rrachire interface functions. The system is configured in a fault tolerant design. It has a fully automatic fail-over capability . If a central processsing unit (CPU) or a portion of nanary fails, the system will automatically reconfigure itself and continue to fully perform its designated functions. 2.2 INPUZ SYSTEM Figure 2.2 shows the schematic diagram for the TSC computer System. Input signals from the control room and other plant locations are taken to the re@ate Input/~t (I/O) cabinets. Signal isolators . are provided in the I/O cabinets so that. no failure on the output side of the I/O cabinets will affect the input signals. In addition to these isolators, all signals coming fran the safety systems are taken after the existing qualified isolators on these systems. The input signals, after going through the isolators, will be converted to binary information on the input cards and then are noltiplexed to the cemputer. Each signal channel has 'ts own Analog/Digital Converter, thus providing a high degree of realiability for the input system. AEP-4

v 2.3 DATA DISPLAY SYSTEM 2.3.1 Technical Su rt Center Room Each D. C. Gook Unit has a dedicated comm-md console located in the Onsite Technical Support Center. Each canmand console is equipped with two color CRP displays and a video hard copier (which can be used to obtain a hard copy of the screen image) . One CRT is dedicated to the PSSD function and the second CRT is a general purpose display. Three satellite stations, each with a color CRT display, are also provided. The satellite stations can be connected to either Cook Unit 1 or Unit 2 TSC Computer System. A shared video hard copier is provided for the three satellite CRTs. We satellite stations are arranged so that visual access from the conuaand station can be maintained while still providing sufficient room to minimize noise and disturbance. Por printing lengthy reports, a hard copy terminal is provided for each Cook Unit Corrputer System 2.3.2 Control Room: Two redundant PSSD display CRTs and two redundant BISI CRTs are provided in each control room. A video hard copier is also provided to obtain hard copy output from the CRT screen image. 2.3.3 Eme~ren. ratin Facilities (EOF): A color CRT terminal, which can access either Cook unit TSC computer, is provided in the Erergency Operating Faciliies. The reste CRT can be used to display all of the displays available on AEP-5

the PSSD, TSC and BISX functions except for the tcp level iconic display of the PSSD function. This iconic display was designed for early recognition of an event by the control rocm operators and therefore is not included in the EOF. AEP-6

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Sensor Signals non-safety Safety syst. syst. signals signals isolators
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solators Process Computer BISI Disrlays 1 I lI L I/O Canine = I/O Ca! ine i I 1 I i PSSD Displays rain A Train D I I CONTROL ROQiI I I I I I I I r
'PSSD Displays I ~BISI Displays L
r TSC Displays I TSC CO>'IPUTCR SYSTE'.1 TECH SUPPORT CE'NTER site I oundary Figure 2.2: TSC Computer System Schematic. TSC BISI PSSD ÃUCLEA'R (non- DATii. iconic) LIxlE< EOF AEP-8
3. ONSITE TECHNICAL SUPPORT CENTER

3. 1 DESI'ASIS:
The Onsite Technical Support Center (OTSC) serves as the focal point for post-accident recovery managerent. As such, it must have the capability to access, display and transmit pertinent plant status information independent of actions in the control room. The Technical Support Center function of the TSC Ccmguter System was designed to satisfy the following requirem nts:
1. Personnel in the OTSC rmst have access to the real time information defining the current status of critical plant systems and functions.
2.,The TSC function aust have the capability to store historical prevent and post~vent data in order to enable a diagnosis and evaluation of the event to determine the extent of any possible plant system damage.
3. The TSC function nust have the capability to access and display plant pazam ters independent of actions in the control zoom.

4. The interface of the TSC system equipment with exisiting plant instrumentation must not result in any degradation of the plant protection system, control roan or other functions.

5. Param tezs to the extent possible should be fran the sana source that is used for control rocm indications to ensure data consistency.

6. The TSC system nust have the capability of interfacing with cannunication equiprrant'or the off site transmission of pertinent plant data.
AEP-9
7. The users est be able to create or edify displays to rraet the needs as conditions may 'dictate.
3.2 INPUZ EETERNIHRTION In order to define the information which must be available in the OTSC,' generic study of critical plant systems and key safety functions (as listed in Table 3.1) was conducted by Westinghouse. This study resulted. in a list of param ters to be nanitoxed by the computer for the Technical Support Center function. This Westinghouse parameter list was reviewed and made Cook Plant specific by AEP. Table 3.2 lists the principal parameters and Table 3.3 lists the basis for input selection. Redundancy and diversity of process indications axe utilized to satisfy cono xns associated with unavailable signals due to sensor failure. Som refinement of the input parameters list may be made after the submittal of this cono ptual design report AFP-10
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3. 3 OTSC OPERATOR INTERFACE The ability of the OTSC to be an effective tool $ n post-eccicfent recovery management is a function of the inputs provided and the abel)ty to present information -in a meaningful and organized manner. As stated previously, the man-machine interface $ s through the use of interactive IJ graphic color CRT displays. The interface functions in the OTSC consist of displays and console functions.
0 The display types available for OTSC personnel use consist of graphic and alphanumeric displays which are both prcformattcd and user construc-tible. Examples of the types of displays available are shown in Figures 3->. 3-2 and 3-3-Figure 3 lis an example of a preformatted system
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status display, g~thering important system and loop parameters onto a single page of display. Figure 3.2 shows morc detailed information on individual parameters such as information on sensor status, current value, and high and low limits.. Figure 3-3 is an example of a graphic trend display showing a time history of related parameters. Highlight-
-ing techniques for indicating parameters or conditions of interest util-ize both color and achromatic means.
By providing a combination of both 'preformetted and user constructible displays the OTSC personne'I are provided with prearranged quickly acces-sible system information and the flexibility to permit the ta',loring of information presentation to meet specific needs as conditions dictate. The specific content of preformatted displays will be determined by analyzing post-accident data requirements in terms of event evaluation, the safety status of the plant, and long-term recovery planning. Dis-plays will also. be designed to reflect plant specific design details.
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Display access is provided both by dedicated functional console push-buttons and standard keyboard entries. Dedicated keys provide access to the most frequently used displays or functions. For other functions access can be either direct by entering shor t codes or by utilizing an instruction function to determine the )dentif)cation code for a display if it is unknown. 5251A AEP-ll
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~ ~ ~ I ~ Other types of information is available through the console keyboard. These consist of fdnctions such as po~nt revie~, logs, post-trip histor-ical data revieg and offsite data transmission. s~ The point rev'iew t functions enable the console operator to 'review plant sensor information. The types of review functions available are: t L. Values of individual points.
2. Points removed from scan.

3. Points removed from limit checking.

4. Points failed under quality checking routines.
$. Points whose'scan frequencies have been changed from ti.e -normal scan
Trequenc ies.
-There are log functions available to the OTSC personnel which can be displayed on CRTs with periodic updates or output onto a hard copy .device such as a line printer. These functions can be preprogranmed and automatically initiated or specified and initiated by console operator input.
The post-trip review function provides the capabil ity to review histor-
-ical data to aid in an event evaluation. This function continuously .stores in memory an updated table of preassigned sensor values for a predefined period. Upon the occurrence of a disturbance (e.g., plant trip) the system continues to store data for a defined time period.
After this period, the entire data record can be reviewed by the OTSC personnel on CRTs and/or output to hard copy devices for permanent record storage purposes. 2-8 O'RC1 A AEP-12
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The offsite data transmi "sion function enables OTSC personnel to'rans-mit plant data to offsite ',ocations via owner supplied comounications systems. The OTSC operator can initiate transmission of data either on a "one-shot" or periodic basis. The transmitted data can be arranged into four edited versions for the specific needs of separate offsite comaunications receivers such as the NRC.
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TABLE S.1 CRITICAL PLANT SYSTEMS/FUNCTIONS
,Reactivity Control Primary System Inventory Core Heat Removal Capabilities Availability and Capacity of Heat Sinks ~ ~
Containment Integrity
; ~ 'Primary System Pressure and Temper ature Availability and Capacity of Alternate Mater Sources I ~ .Availability and Operability of Critical Support Systems Radioactivity Control 2-10 AEP 14
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~ ~ I~ ~~ ~ Table 3.2 TSC Param ters List Variables Min. No of Si ls
-BCS hot leg temp 0-700 deg F -RCS cold leg temp 0-700 deg F -BCS pressure 2 0-3000 psig -Beactor water level 0-100 % -BCS boron concentration 0-5000 ppm -Pressurizer water level 0-100 8 -Steam generator level Wide range 0-100 8 Narrow range 0-100 8 -Steam line pressure 0-1400 psig ~ntainm nt pressure -~36 psig -Containment water level 589'-'599'lev.
599'-614'lev.
-RNST water level 0-100 8 condensate storage tank level 2 0-100 0 -Boric acid tank level 0-100 % -Aux feed water flow 0-250 Klbs/hr ~Rain feed water flaw 0-5000 Klbs/hr -High head injection flaw 0-200 gpm AEP-15
Table 3.2 TSC Pazmn ters List Variables Min. No of Si ls ~Ran ~ -Zaw head injection flow 4 0-5500 gpm -Core exit temperature 16 0-2500 deg F component cooling water flow 2 0-10000 gpm -Carponent cooling water temp. 2 32-200 deg F ~ntainnant hydzogen concent. 2 0-30 % -Containment tempezatuze 0-100 deg F ~utzon flux 0-120 0 power -Control zod position 53 'Full in or mt -Primary system relief & 4 Closed-not closed safety valves -Sec. syst. relief valves 4 Closed-not closed -Containment isolation valves 139 Closed-not closed -PZR relief tank pressure 1 0-100 psig -PZR zelief tank level 0-100 8 -PZR relief tarik temp. 1 50-350 deg F 1 -RCS degree of subcooling N/A 200 sub-5 super -Accumulator level 0-100 8 -Accunulator pressure 0-700 psig -Accumulator isolation valves Closed-not closed -Aux building sump level 0-flocd level -RHR system flow 0-7000 gpm AEP-16
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' I Table 3.2 TSC Paraneters List Variables Min. No of Si als -RHR heat eIc. outlet temp. 0-400 deg F -Boric acid charging flaw 0-10 gpm -RCS let dawn flaw 0-200 gpn -RCS make-up flaw 0-200 gpn -Em rg. ventilation ~r closed-not closed -Status af standby power Energized or not -High radioactivity liquid 0-100 8 tank level -Radioactive gas decay tk press 4 0-150 psig -Reactor Coolant Punps status 4 0-1200 anps -PZR heater bank status 0-200 arrps ~meteorology Wind direction 0-360 deg Wind speed 0-100 miles/hr Atm. delta temp. 0-50 deg F -Radiation 2 Containm nt area radiation 1 .1-10E4 mR/hr Containm nt radio gas 1 10-10E6 cpm Containtnent air particulate 1 10-10E6 cpn Unit Vent radio gas 10-10E6 cpn Unit Vent iodine 10-10E6 cpm AEP-17
Table 3.2 I III Variables Min No of Si ls I -Radiation (continued) Steam gen. blow down 10-10E6 cpn Condenser air ejector . 1-10E4 mR/hr Cooling water East 10-10E6 cpm Cooling water West 10-10E6 cpn Service water East 10-10E6 cpn Sexvim water West 10-10E6 cpn Waste liquid off-gas 10-10E6 cpm Waste gas decay tarik 10-10E6 cpm Control rocm area . 1-10E4 mR/hr Spent fuel area . 1-10E4 mR/hr Charging pp rocm area . 1-10E4 mR/hr Note 1: Degree of subcooling will be independently calculated by the TSC ccmputer Note 2: The radiation signals listed above are signals from the existing radiation detectors. AEP is in the orocess of ingle'ting a new Radiation Monitor System at Cock Units 1 and 2, and will transmit the required radiation signals to the TSC ccmputer from this new Radiation Monitor System AEP-18
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TABLE 2-3 TSC INSTRUMENT BASIS PARAMETER INITIAL EVENl OIAGNOSIS" BASIS I& I tran I 5251A
TABLE 2-3 (Continued) TSC INSTRUMENT BASIS PARAMETER INITIAL EVENT DIAGNOSIS* BASIS QJ I ~ CO 5251A
T"BLE 2-3 (Continued) TSC INSTRUHENT BASIS PARfit lETER INITIAL EVENT DIAGNOSIS* BASIS 525]A
TABLE P-3 (ContInuerl) TSC INSTRUMENT OASIS PARAMETER INITIAL EVENT OIAGNOSIS* OASIS
TABLE 2-3 (Continued) TSC INSTRUHFNT BASIS PARAHETER INITIAL EVENT DIAGNOSIS* BASIS 5251A
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TABLE g. 3 (CogtIpoed) TSC INSTRUMENT BASIS PARAMETER INITIAL EVE NT D JAGNOS IS* BASIS
TABLE 3 3 (Continued) TSC INSTRUMENT BASIS PARAt 1ETER INITIAL EVENT DIAGNOSIS* BASIS (b,c) I Vl 'i 5251A
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TABLE 2-3 (Continued) TSC INSTRU)hENT BASIS PARAihETER INITIAL EVENT DIAGNOSIS* BASIS (b,c) 5251A
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~ ~ Systems Status - Reactor Coolant System Loop 1 Loop 2 Loop 3 Loop 4 T average ('F) 595.2 595.2 595.2 595.2 Overpower DT ('/oPWR), 110.0 110.0 110.0 110.0 Overtemp, ZD ('/0PVIR) 110.0 110.0 110.0 110.0 Cold leg temp. (narrow range) ('F) 559.8 559.8 559.8 559.8 Hot leg temp. (narrow range) )'F) 624.0 '24.0 624.0 624.0 Reactor coolant flow (%) 100.0 100.0 100.0 100.0 Reactor coolant pressure - WR (PSIG) 2250.0 2250.0 2250.0 2250.0 Pressurizer pressure (PSIA) 2250.0 Pressurizer vapor temp. ('F) 563.8 Pressurizer liquid temp. ('F) 565.2 Pressurizer relief tank pressure (PSIG) 1.5 Pressurizer relief tank level (%) 77.6 Pressurizer relief tank temp. ('F) 110.3 Pressurizer safety relief temp. ('F) 120.0 Figure 3. 1 System Status Display at Qnsite 1 echnical Support Center (Example) AEP-27
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Parameter Summary Point Description Yalue Range Units Status TO400 RCS Loop 1 Hot Leg T 593.4 0:700 . DEGF Normal TO406 RCS Loop 1 Cold Leg T 547.2 0:700 DEGF Normai PO480 RCS Pressure 2234.1 0:3000 PSIG Normal
'LO421 Stm Gen 2 Narrow Range Level 39.1 0:100 PC Low PO549 Steamline Pressure 893,0 0:1100 PSlG Normal LO103 RWST Level 100.0 0:100 PC Normal LO114 Boric Acid Tank Level 98.8 0:100 PC Normal LO119 Condensate Storage Tank Level 56 4 0:100 PC Normal LO947 Containment Bldg.'Vater Level 3.3 0:100 PC High Figure 3. 2, Parameter information Display at Onsite Technical Support Center (Example)
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700 RCS COLD LEG TEMP (oF)
'100 700 RCS HOT LEG TEIVIP (oF) 100 100 PRZR LEVEL (~o) 40 2500 PRZR PRESSUR E (PSIG) 1900 2 4 6 8 10 12 14 16 18 20 TIME (SECONDS)
Figure 3. 3Graphic Display at Onsite Technical Support Center (Example) E AEP-29
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4.0 PLANT SAFETY STATUS DIS?LQY 4.> PURPOSE The function of the Plant Safety Status Display (PSSD) is to present a succinct account of the overall plant safety status to the control room operator (or supervisor). The entire data base should be available to the operator arranged in a format that will enhance his response to events and the diagnoses of the cause of the event. Because the 'PSSD serves as an important interface between the plant process and the operator, the information presentation should be defined in terms of parameters and logic supportive of defined operating procedures for dealing with abnormal events. 4e2 INPUT DETERMINATION In order to dete'rmine thc required operational modes for the PSSO (b,c,e) Because of the fact that gi (b,c,e) {b,c,e) 4-1 54".5A
The parameters available for The role for which the PSSO provides L J is as follows: (b,c,e) ~, 4-2 AEP-31 5435A
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(b,c,e) By addressing L (b,c,e) (b,c,e) for the (b,c,e) In defining the inputs PSSD, L 1as fo11ows:. ~ c (b,c,e) In response to the/ (b,c,e) O. 4-3 PEP-32 v's.I vvg QC.lI
4e 3 MAN-MACHINE INTERFACE The PSSD system will process the defined input data set of plant param-(a,b,c) eters at Q Q and generate displays for redundant PSSD (a,c) dedicated CRTs located in the control room. In order to achieve an effective man-machine interface, the display system must be designed to provide a logical and human engineered dis- ~ . play structure and selection process in a manner which supports defined roles in which the operator is expected to per form during an abnormal occurrence. The role of the control room operator in/~ depicted in Figure 4-1. The display system structure should be defined such that it L gare defined as follows: 2 ~ 4 4 AEP.-3 3 54~=A
~ W s ~ (b,c) The display structure shown in Figure 4-2 Q (a,c,f) 4-5 AEP-3 4
(a,c,f) A major problem associated with the man-machine interface is the (a,c, f) figure (a,c,f) 4-3 is an illustration of the display. ~ (a,c,f) Figures'-4 and 4-5 are preliminary versions ofl for two sample events: Primary to Secondary Coolant System Leak and Primary Coolant System Leak to Containment. The parameters chosen for the displays were chosen to (a,c,f) (a,c,f) 4-6 AEP-3 5 5435A
The inforaatlon atL 4-7 5435A AFP-36
TABLE 4-1 PLANT SAFETY STATUS OISPLAY - SAFETY GOALS - TERMINATE MOOE TRANSIENTS 4-8 AEP-37 i435A
TABLE 4-2 PLANT SAFETY STATUS DISPLAY - SAFETY GOALS - MITIGATE MODE TRANSIENTS 5435A mP-38
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TASLE 4-2 (Continued) PLANT SAFETY STATUS DISPLAY - SAFETY GOALS - MITIGATE t',ODE TRANSIENTS 4-10 AEP-39 5435A
TABLE 4-3
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PLANT SAFETY STATUS DISPLAY TERi~tINATE YCOE PARAllETERS {b,c,e) 4-11 AEP-40
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TABLE 4& PLANT SAFETY STATUS P~SPLAY NITTGATE ."POE PARAvETHS 4-12
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~ ~ ~ I q( Figure 4-2. Display Structure of Plant Status Display AEP-43
~ ~ ~ ~ (a,c,f) Figure 4-3. Sample Display Plant Safety Status Display AEP-44
Figure 4.4. Sample Plant Safety Status Display Temiinate Mode Primary to Secondary Coolant System Leak (SG Tube Leak) AEP-45
(a,c,f) Figure 4-5. Sample Plant Safety Status Display Mitigate Mode Primary Coolant System Leak to Containment AEP-46
.5 ~ 0.. BYPASSE0 ANO INOPERABLE STATUS INOICATIGN FOR PLANT SAFETY SYSTEMS
5. 1 PURPOSE The purpose of the Bypassed and Inoperable Status Indication (BISI) system is to provide the control room operator with a continuous systems
.level indication of a bypassed or inoperable condition for the systems compr ising the engineered saf ety features. The system consi ders the actual status of individual components including systems level bypasses and control room operator entered inputs for components removed from service.
5. 2 INPUT OETERMINATION Bypassed and inoperable status indication is provided for the systems comprising the engineered safety features and their critical support systems. These systems are identified in Table 5. l. This table also identifies the types of components for which monitoring is required, the approximate number of each type of component, and the type of status
.information needed. This list is generic in nature and will be revised to meet individual plant specific designs. In the evaluation of system inputs, the components in each system are . considered in the light of being in a proper state to perform or supoort the operation of a safety function. The systems level bypass functions that must also be considered are listed in Table 5 . 2 . In addition to automatically monitored inputs, the system also considers the effect of component or system out of service inputs manually entered by the control room operator .
5. 3 MAN-MACHINE INTERFACE The interface between the operator and this system is provided by redun-dant CRT displays and keyboard consoles located in the control room.
Personnel located in the Onsite Technical Support Center will also be
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able to access the same information. The 8IGI utilizes a structured display hierarchy for the ooerator interface. The display hierarchy is shown in Figure 5 . 1. The primary display, an example of which is shown in Figure 5.2i con-
,tains the following information for each of the systems comprising the engineered safety features:
1. Bypassed or inoperable statu" indication for each affected subsystem on either a systems level and/or train level basis.

2. Identification of whether the condition is due to the inoperable status of a component or auxiliary support such as cooling water, power supply, tc.
Other levels of displays such as shown in Figure 5. 3 provide supporting information on individual components within each subsystem and support (a,c,f) system. Whenever the status of a system becomes inoperable or bypassed, the (a . r) AEP-48
3-3 AEP-49
~ ~' I
~ ~ ~ ~ TABLE s.a BYPASSED AtsD IttOPERABLE STATUS IttDICATION-COMPOttettT IttPUTS
~Sstem Com onents Statvs (b,c)
Emergency core cooling Valves Open/Shut Pumps Operable
-.Process High/Low, etc.
(level,pr essur e)
.Auxiliary feedwater Valves Open/Shut . Pumps . Operable a i ~ . Process .High/Low, etc.
Containment spray'alves Open/Shut Pumps. Operable
'rocess High/Low, etc.
Containment isolation Valves Open/Shut Auxiliary power system Breakers Open/Closed/Out
'enerators Operable Voltages High/Low Containment ventilation Valves Open/Shut Motors Operable Containment hydrogen Valves Open/Shut recombiners Motors Operable Component cooIing Valves Open/Shut Pumps Operable Service water Valves Open/Shut Pumps Operable 3-4 AEP-50
~ ~ ~ ~ TABLE 5. 2 BYPASSED AND INOPERABLE STATUS INDICATION-SYSTEM LEVEL BYPASS FUNCTIONS Safety injection
- 'ow pressurizer pressure Low steamline pressure Manual reset Steamline isolation Steam dump interlock Steam generator blowdown isolation 3-5 AEP-5 j.
5251A
~ ~ 1 ~ I
Figure 5- j Display Structure Bypassed and Inoperab!e Status Indica;ion AEP-52
Figure 5. 2 Primary Disolay Bypassed and tnoperaole Status indication AEP-53
l ~ I
~ ~ ~ ~ (a,c,f) Figure 5.3 Secondary Display Bypassed and Inoperable Status Information AEP-54
~ )
6. TSC &STD~>'lATlCH
\ I As described in Section 2, Past of the input signals to the TSC cavputer are ~en frctn the existing instruments which also provide signals for 12m Control Rccm indicators. This approach will provide consistent data in both &~ contxol zoom, Onsite Technical S~rt Center and the EOP. TI ~ input signals to the TSC computer therefore have sare hign quality, accuracy and xeliability as the contxol zootn signal. Transfozner isolators are provided for all analog input signals and mtical isolatozs are provided for all digital input signals.
~w addition, all signals fxan the Faactor Protection Channels aze ~n after the n~ sting safety grade isolatozs. C~oze, M interfacing of t>a TSC cmgi r system to the existing plant instzmntation will not result in ary degradation of th contxol zocm, protection sys"w, controls oz oi9mr plant functions. AEP-55
'I ( ~ ~ ~ ~ ~ I ~ j
7. TSC POWER SUPPLY SYSTEMS 7.1 POWER TO THE TSC CGMPUZER SYK'EM:
will be (UPS). The power xequixeaants satisfied though the This UPS system use of an will provide of the the TSC unintex~ptible TSC
~
Computer System supply system cxmputexs and peripheral equipment with a high quality, transient free power source. 7.1.1 THE UPS SYSTEM: Figure 7.1 shows a one-line diagram (schematic) for the UPS system. The system consists of redundant battery chargers, battery, static invertexs, and static transfer switches. Under normal conditions, the battery chaxger converts AC to DC and supplies it. to the inverter. The battery charger also keeps the battery at full charge. The inverter converts the DC to AC in order to supply the load requirements of the TSC ~ters and their peripheral equipHBIlto 7 1.2 CONSEQUMZS OF POWER SUPPLY INTERRUPTION: If there is a power reduction (dip or degradation) or loss (failure) of the AC power source, the UPS battexy becones the prirmxy source of DC to the inverter, rather than the battery charger which has lost its normal source of AC power supply. The battery will be sized to supply the inverter load xe'quixem nt for a period of 30 minutes. This allows a sufficient time interval in Mich a diesel generator (badmp AC source) can be made available to provide power to the inverter. In the unlikely event of loss or AEP-56
TSC POWER SUPPLY SYSTEM (CONCEPTUAL DESIGN) EMERGENCY SOURCE NORMAL SOURCE BACK-UP SOURCE INDEPENDENT INDEPENDENT INDEPENDENT 600 VOLT BUS 600 VOLT BUS 600 VOLT BUS M.C. C. M.C.C. BREAKER BREAKER 225A 225A M.C,C. AUTOMATIC BREAKER TRANSFER 225A S W ITCH 260A I 600 ~700 AMP~ 700 AMP 75KVA BATTERY BATTERY I20 CHARGER
~~(ALTERNATE) CHARGER L
BATTERY 927A 40KVA 40KVA INVERTER INVERTER STATIC STATIC SWITCH SWITCH
'IGURE 7.1 UNIT W I UNIT W 2 TSC TSC COMPUTER 8 COMPUTER 8 P ER IP HERA LS PERIPHERALS 6/I'/8I I
AEP 56 a
unavailability of both the normal and backup AC sources, the static switch will be used for transfer, if necessary, to the em rgency AC source e 7.2 KNER TO THE TSC COMPLEX: Standard balance-of-plant (BOP) sources will provide the TSC with power for lighting and convenience receptacles. For additional protection, the lighting fixtures are provided with battery packs for continued operation in the event of loss of the BOP power supply. The HVAC equiparant will.be supplied frcm an Essential Services System bus (AC source) . AEP-57
0
~ ll
8. TSC AND EOF FACTIONS 8.1 TASK FUNCTIONS PERFORMED BY INDIVIDUALS IN THE TSC:
The emergency functions/tasks perfozmed by . individuals required to report to the TSC are described kp the follawing: S.l.l RADIATION MOND)RING
.Coordinate activities of field assessment teams.
Receive Data from personnel in the field. Provide RADIATION MONITORING MTA to appropriate personnel. Dosimetry Contxol. 8.1.2 DOSE ASSESBKNT: Receive data from comtlUnications personnel on Radiation Monitoring and Meteorological conditions. Receive Radiation Monitoring Data from Radiation Monitoring Director. Perform Ebse Assesarent calculations. Provide recommended protective actions, as neceessary. Assist in classification of event. Place Contzol Boom Data in format useful to off-site agencies. 8.1. 3 COMMUNICATIONS Receive event data fran Contzol Room. Place event data in format useful for Dose Assessm nt personnel, Plant Status Evaluation personnel, and Off-site Agencies. AEP-58
Ccmnunicate event conditions to the Berrien County Sherif 's f Department; Michigan State Police; the Joint Public Information Center; AEP Service Corporation; and Industry Support Groups. 8.1.4 TECHNICAL SUPPORT Provide technical support to Operations personnel in areas such such as Core Analysis, Chemical Control, +cle Evaluation, and instrumentation.
- Provide Independent Evaluation of the Safety status of the Unit.
8.1.5 MANAGEMENT SUPPORT Provide for availability of site personnel as needed. Provide direction and priorities for site personnel activities.
- Provide evaluation of Emergency naasures to be taken on-site and off-site.
8 2 BKBCZNCY FUNCTIONS PERFORMED IN THE TSC/EOF FOR EACH EMEBGENCY CLASS: 8.2.1 QKSUAL EVENT: Plant conditions requiring declaration of an "Unusual Event" are not expected to >+quire activation of the Technical Support Center or Emergency Operations Facility. The Emergency functions previously described will be delegated/coordinated from the Control AEP-59
~ 4 a~ 4~ g ~
e ~ .>:
8.2.2 ALERT
Plant conditions requiring declaration of an "Alert" vary in severity level frcm the upper bounds of "thusual Event" to tlat lower bounds of a "Site Emergency". An "Alert" classification thexefore, may require performance of portions of the functions, described above in Section 8.1, in the Technical Support Center and the Control Boom, or it may require performance of all of the emergency functions in the Technical Support Center and the Etta~cy ~tions facility. The degree of activation of the TSC/EOF is a function of time as well as of event severity. At the tirade the event occurs all emergency functions will be perforned in the Control Rccm, with first priority for operator actions given to event mitigation. Apizcodmately 1 hour after event occurrence the required emergency functions will be divided between the Control Room and the TSC. Zf the event is of low severity within the "Alert" category, a majority of these functions as applicable to the event. within the "Alert" category, will be Zf the it is
~
performed expected in the Contxol Boom, is of high severity that the majority of these functions will be performed in the TSC, as applicable to the event. PAP-6 0
~ k
+~ ~ 's >.a
If the event continues for a long period of tim, such as 24 hours or naze, response group arrivals at the site will requize full activation of the TSC and a partial activation of the EOF, independent of the relative severity of the "Alert" event. All applicable emergency functions will be performed in the TSC with the exception of dose assesmant, which may shift to the EOF. 8.2.3 SITE AND GRIEF BKRGEKZ A "Site Emergency" will require full activation of the TSC. Except for the dose assessment function, which may shift to the EOF, all applicable emergency functions of Section 8.1 will be performed in the TSC. A "Site Emergency" is not expected to occur instantarmusly; however, should this occur, the TSC will be activated, staffed and assume emergency functions frcm the Control Rocm, within 1 hour of event occurrence. The EOF is expected to be activated within 4 to 6 houzs and it will assuna the dose assessment function from the TSC. A "General Emezgency" will require full activation of the TSC and EOF and the functions performed by site personnel assigned in these facilities is expected to be identical to those functions perforrred for a "Site Em zgency." AEP-61
8.3 FACTIONS OF INDIVZDGKS REPORTIh6 TO THE EOF: The emergency function/task perforrred by indviduals required to report to the EOF are chscribed in detail in the DCCNP Enargency plan Chapter 12.3.3.3 and are generalized below by the following four catecpries: Coordination of Off-site Radiological Monitoring and Dose Assessment.
- Technical ~rt of Plant Recovery Operations.
Managenant Support of Recovery Operations.
- Communication with Offsite Agencies.
AEP-62
Ah> 4 ~
~ l ~.
9. TSC RECORDS AND DATA AVAILABILITY It is necessary to make available in the Technical Support Center the reference material .and data source material needed to'ake a technical evaluation of an accident or em rgency situation. Therefore, up-~te plant specific docum nts and general technical references needed to implenent this function will be maintained in the Technical Support Center.
9.1 CONTROLLED PLANT SPECIFIC REFEREKZ MATERIAL: For plant specific reference, the follawing controlled material will be kept in the TSC: Technical Sepcif ications. Abnormal and Emergency Operating Procedures. Detailed eleaantary electrical diagrams, and detailed flaw diagrams. Included with this information are plant arran~t diagrams showing component locations. Contour area map with population distribution and overlays for'lum evaluation. Donald C. Gxk Nuclear Plant Em rgency Plan with procedures. System Descriptions. Precautions, Limitations and Setpoints Plant. Technical Data Bock containing curves for reactivity control, rod worth, RSC temperature and pressure limits and secondary plant performance. AEP-63
The above material will be controlled by the Donald C. Cock Plant docum nt control system which is governed by Plant Manager Instruction 2030, entitled Eocurrent Control. 9.2 UNCONTROLLED INEORMATIQN AND TZXZiNlCAL REFERENCE MATERIALS: In addition, other plant information which is useful will be present in the TSC: Pump GIld fan performance c~lBs ~ Final Safety Analysis Report. Annunciator Layouts. Tank Volum /level curves. Apprcpriate Plant Manager Procedures. Reference copies of miscellaneous em rgency procedures; and Unit Vent Emergency Release Level Determination and Secondary System Em rgency Release Determination guides. General technical reference materials will also be available, such as: Steam tables. Chart of Nuclides. Standard Handbock for Electrical Engineers. Handhxk of Chemistry and Physics. Standard Handbook for Mechnaical Engineers. Ghermacbpmmics Handbock. Nuclear Reactor Engineering Handbook. Radiological Health Handbxk. Instrument Engineers Handbock. Pump Handbook. AEP-64
The above~ntioned items will be maintained in the TSC readily available and other materials dec@ed necessary may be added in the future at arp tiae.
9. 3 OZHER DATA, RECORDS, AND INFORMATION:
The following references axe immediately available to the personnel in the Contxol Boom: Abnormal Operating Procedures The following reports and information are available in the plant library. Plant Operating Recoxds Plant Nuclear Safety Review Canmttee Beports Vendor manuals, and cmponent level drawing and sketches State of Michigan and local 'em rgency preparedness plans. The plant library is located in the office building, Mich is in close proximity to the TSC and Contxol Rooms. AEP-65
t t f rb 1
~ '
Attachment C to AEP:NRC:0745H Responses to Questions on Unit l Cycle 8 Thermal Hydraulics (Non-Proprietary)
QUESTIONS'1 throu h 3: What is the basis for using the WRB-1 correlation for 15xl5 OFA? What is the basis for the DNBR limit of 1.17 for WRB-l applied to 15xl 5 OFA? Are there any critical heat flux data?
RESPONSE
The WRB-1 CHF. correlation is based enti'rely on rod bundle data and has been shown to provide a significant improvement in DNB predictive capability for Westinghouse fuel designs with type "R" mixing vane grids. C The NRC has recognized this increased accuracy and concurred that a 95/95 limit DNBR of 1.17 is appropriate for 12 ft and 14 ft 17x17 standard and optimized fuel assemblies, and 12 ft 15xl5 standard fuel assemblies with the type "R" mixing vane grid'Ref. 1 and 2). Based on the semi-empirical nature of the correlation, the NRC has imposed restrictions on its appli-cability to other PWR designs. Specifically, the Safety Evaluation Report stated that, "The correlation should not be applied to any PWR geometry which has not been specifically tested or which has not been bracketed by the test data. The important parameters to which this applies are: rod size, rod pitch, heated length, mixing vane design and grid spacing." T e 15xl5 optimized design is virtually identical to the 15xl5 R-grid design in that the I.
As will be discussed be]ow, similar scaling techniques have been used for designing the 1?xl7 OFA and 14xl4 OFA grids, and DNB testing has shown that the MRB-1 correlation correctly predicts the performance of those designs without modifications. Based on the previous success of this grid scaling technique (as demonstrated by the 17xl7 OFA and 14xl4 OFA DNB test results) and the similarity of the 15x15 OFA and R-grid geom tries, use of the MRB-1 CHF correlation with a de-sign limit of 1.17 is justified. for the'5x15 OFA design. 17xl7 OFA DNB Test Results Geom trically the'?x'l7 OFA design differs from the standard 17xl7 R-grid design in that
1) The fuel rod diameter was reduced from 0'.374 inch to
[ g inch. ( a,e)
2) The Zircaloy type "R" grid is [
] than the fnconel type "R" grid which has previously (a., c) 'een DNB tested.
0 In order to minimize the effect of the grid dimensional'changes on DNB performance, special care was taken to preserve the important type "R" mixing vane characteristics. [ (CL~ Q)
+ .] DNB testing of the 17xl7 OFA geometry demonstrated the success of this scaling approach the MRB-1 correlation predicted the data well without any modifications, using the..same performance factor as was used for the 1?xl? standard fuel. Re-peatability studies (Ref. 3) have shown that the accuracy of the MRB-1 correlation is essentially identical for the 1?xl? OFA and standard geometries, indicating that no additional component of variance is introduced by the grid dimensional changes. In other words, the correlation correctly accounted for the equivalent diameter effects and the scaling approach correctly accounted for 'the grid dimensional changes..
14xl4 OFA DHB'Test Results The 14x14 optimized geometry differs from the standard geom try in that:
1) The fuel rod diameter. has been-reduced from 0.422 inch to
] inch.
2) The Zi.rcaloy type "R" grid is [ )+
than the .?nconel type "R" grid which had previously been DNB tested. A CHF<est series. of the 14xl4 OFA'typical cell geometry has been perform d'to verify that the MRB-1 correlation correctly'redicts the effect on CHF of the equivalent diameter change, and that the grid scaling approach introduces no additional component of variance.. As will be discussed below, the results ir-icate that the WRB-1 correlation predicts the 14xl4 OFA data with essent;ally the same accuracy as for the geom try from which it was scaled. Test Fhcilitieh The test facilities and testing procedures used f'r the 14x14 OFA CHF tests were the same as those described in References' and'5. The test. section was similar to the 0.422 inch rod bundle described in Reference 4. except that the mixing vane grid dimensions were modified slightly in order to accommodate the new rod diameter and the change from Inconel to Xircaloy. The modified grid design has retained the type "R" grid features. Figure 1 shows a sketch of the 14xl4 GFA 'typical cell test bundle cross section. The axia1 locations of the grids .an'd thermocouples are shown in Fi'gure 2, and Figure 3 sh'ows the cosine axial power distribution used for the tests.
CHF Data Evaluations The data were reduced using the THING subchannel code, in the same manner as described previously in References 4 and 5. The WRB-.l correlation of Reference 6 was used to predict the critical heat flux. The performance factor used was the same as that employed for the 0.422 'inch data eval-uations [ ], since the mixing vane. grid size was [
+ ] As discussed above, this approach had previously worked quite well with the 17x17 OFA CHF data.
The .results of the data reduction are shown in Table 1. The average measured-to predi"c'ted critical heat flux ratio for the data set is [ ] with a 5,c
+
sample eviation oof [ standard deviation ] . These values were compared to those .(44 from the 0.422 inch rod bundle tests with 26 inc inch gri'd spacing', ththe geometry from the original WRB-1 R-grid database which is closest to the 14xl4 OFA geometry. As shown in Table 2 the agreement is excellent, indicating that the WRB-I correlation correctly accounts for the geometry changes and that the choice of performance factor is appropriate. Also given in Table 2 is a comparison of the 17x17 standard and OFA DNB statistics. It is apparent that t¹'RB-1 correlation'0 ability to predict CHF is es t'ly identical for standard and OFA fuel designs, T-tests and F-tests have been perform d for each of these standard/OFA data set pairs in order to evaluate the effect of the geometry changes on .the accuracy of the WRB-I correlation. Table 3 shows the results of these tests. It can be seen that the hypothesis that the WRB-1 correlation pre-the-{.'~+ diets the DNB behavior of the OFA geometries with the same accuracy as the standard R-grid geometries cannot be rejected at a 5l significance level, with the exception of ;] comparison. For that comparison the OFA data had an appreciably lower variance . A smaller variance is indicative of better correlation accuracy, so failure of the F--test e is no reason for concern. Therefore, the results of these tests indicate that no additional component of variance is introduced by the grid dimensional changes.
FIGURE 1 FIGURE 2 0 AXIAL GRID AND CHF DETECTOR LOCATIONS, 14X14 OPTIMIZED CHF TEST SECTION
Z,,AXIAL DISTAHCE FROM BEGIHHIHG OF HEATED LEHQTll '(.IHCIIES) Flgura 3 . Axial Heat Flux Distribution
hatt TABLE 1 - CHF TEST RESUt.fS PgR 14x14 OFA TYRICAL CELL USlt(G j(RD-I CORRELATION tht t tt tt I t St It tthtt the t t t I t tt tet tt I I tI I tt I t the I tt he tht ttI I tt t thth tt tt t thhhl't thht t tt tt t t t t t I t t I i tt I It t t t t t '. INLET INLFT Lt(LET l(ASS . );OCAL LOCAL RE4T FLUR M/P ~ ELEVATIJH FRO)l ?HLEf RVN PRFSSURE TEMP VELOCITY QUAL(TY (XlOE6 BTU/HR-.SQFT) CllF (lACI(ES ) NO ~ .(PSIA) ( F) (X1066 LB}l/HR-SQFT) (P) l PRED ~I (sfRB-1 PREO NEAS <<Q, c1 htPt htt tt I t thttt tttttt ttt tt t t tht tt t I tII t tt tt I t tt+t tt ttht tI't t tIL I't I't tt tttI'tttht tI'thtt ttht t At tt t tt I t It t I t t t AREAS ~ ) ~ I' I g,c) M22%2 1 0599 M2243 1,1582 M22% 0. 1.0593 MZZ%5 1.1O64 M22%6 1.1149 M2247 1'.1483 M2Z's 8 1.0943 M2209 1.1320 M2250 1:1482 M2251 1.1118 M2252 MZZ53 M2250
'. ~ 9741 lo36 1.0983 M2255 1.0939 M2256 1.0844 M2257 1.0783 M2258 1.1O)8 M2Z59 . 9514 M2260 1.1338 'M2Z61 1.0550 MZZ62 .8592 M2263 .9104 MZZ6% . 9466 M2265 1.O739 M2266 1.0461 a2267 .9942 M2268 l.ol54 M2269 1. 0146
~ M2270 1.0065 M2271 . 9615 ttttttttttthtttttttttttIttttttttttttthhtIthttftIhthttttttihfttttttlthttttttthtthhttttttttttttttttttthh+ttttt
TAGLE 1 (CONTINUED) CHF TEST RESULTS FOR. 14xl4 PF/) TYPICAL CELL et t t t t t et et'ti et et et tee ittttiittteetrtititteittt tti tet t t tttteetti'tt tet et et t tet t tet tt eet tet te etc te et et et tet tet Balll/HR IHL ET IHL ET INLET l'ASS LOi AL L3".AL 'lE AT FLUX, H/P ELENATIJN FROh INLET RUN PRESSURE TEMP VELOC?TY QUALI1'r (X10E6 SQFTl CHF (INCHES l (PSIAl l (X10ib LBH/HR-SQFTl (P l (MRG-1 l HJ . ( F tttttttittitttttttttttttettttttttetettettettt
~
ettttttrtti tttttittitti ttttt ttttttti HE AS ~) PREO ~ P RED ~ HEAS etetttttt tttetett tttttete
~I +g,c) +b,c M2272 1.1401 M2273 'I . 0612 M2270 .9741 MZ31% .9314 M2315 1.0735 M2316 1.1411 M2317 1.1465 tttttttitteitttttittttttttttttteeittttt titetttteteeeetii ti tttttieeet tt ttttiiettiteteeitieitttttstettttteeett +'LiC L II0 FT Run O.n.: t.<OO IS]
DE ~ 5800 IH ZIRC SPRlicG HV GRIOS 26 IH SPACING RODS 1004 INhrx RUO/OBITER ROO POMER ~ 1 ~ 1765 1Z RODS 05s
TABLE 2 STATISTICAL COMPARISON OF STANDARD AND OPTIMIZED FUEL CHF RESULTS USING THE WRB-I CORRELATION
TABLE 3 F-test and t;test Results for Standard/OFA Data Set Pairs in, Table 2
For these tests the 0.422 inch rod DNB data sets have been grouped
REFERENCES
1) Letter, D. F. Ross, Jr. (NRC) to D. B. Vassallo (NRC),
Subject:
Topical Report Evaluation for WCAP-8762, April 10,, 1978.
2) Letter, R. L. Tedesco (NRC) to T. M. Anderson (Westinghouse),
Subject:
Acceptance for Referencing Topical Report WCAP-9401 (P)/WCAP-9402 (NP), Hay 7, 1981.
3) Beaumont,. M. D., Skaritka, J., "Verification Testing and Analyses of the 17xl7 Optimized Fuel Assembly," MCAP-9401, March 1979.

4) K. M. Hill, F. E. Motley, F. F. Cadek, A. H. Menzel, "Effect of 17xl7 Fuel Assembly Geom try on DNB," WCAP-8926-P-A (Westinghouse Proprietary) and WCAP-8297-A (Non-proprietary), February 1975.

5) F. E. Hotley, A. H. Menzel, F. F. Cadek, "Critical Heat Flux Testing of 17xl7,Fuel Assembly Geometry with 22-inch Grid Spacing," WCAP-8536, (Westinghouse Proprietary) and MCAP-8537 (Non-proprietary), May 1975.

6) Motley, F. E., Hill, K. M., Cadek, F. F., Shefcheck, J. J:, "New C Westinghouse Correlation MRB-1 for Predicting Critical Heat Flux in Rod Bundles with Mixing Vane Grids," WCAP-8762, (Westinghouse Proprietary),
July 19?6.
QUESTION 4: What is the reason for using two critical heat flux correlations in the same core?
RESPONSE
The Exxon fuel currently in the D.C. Cook Unitni 1 core was originally licensed with the W-3 critical heat flux correlation (
Reference:
0 C ~ ~ Cook Unit 1 Final Safety Analysis Report). Therefore, the Westinghouse analyses wf the Exxon fuel during the transition cycles have also utilized the W-3 correlation. The WRB-1 critical heat flux correlation was developed from a large body of Westinghouse mixing vane grid rod bundle CHF data, and has been shown to predict CHF for fuel designs with'he type "R" gri t g d wi th better b accuracy than previous correlations. The WRB-1 correlation was, therefore, selected for analyses a of the Westinghouse 15x15 optimized ' fuel d esign,' ic h uses mixing vane grids of the type "R" design. Further justification for the use of the WRB-1 correlation for the 15xl5 OFA design is provided in the responses to questions 1 through 3.
HRC VESTION HO. 5 A 5% DNBR penalty for the transition mixed core is used in this reload (D.C. Cook Unit 1 Cycle 8) as a result of analysis using the same methods as applied for the 17xl7 OFA-and 17x17 LOPAR cores. Provide your analyses and results.
RESPONSE
Attached are results of the analyses which were performed in order to calculate the D.C; Cook Unit 1 Cycle 8 transition core DNBR penalty of 5%.
TADI.E . l A IALYSES HADE TO JIIS1 I FY TDAHSI I IDII 000E HETIIODS" A~la I Pnwe> Coul' gura t I on Pressuro In I o t Taupe rn turo Power ,Dlstrlbutlon Ilull (Figure Ilo.) (ps la) ('F) ($ oF 16.8II N(t/Assy) ($ ol'gIII) o ~
/
00 gpm/assy) (Figure IIo, ) 1 2 3 II 5 6 1 9 IO 11 ~ 12 13 I II 15 16 IT 18 19 20
~
2l I 23 I 211 25
TABLE 2. - RESULTS OF TRANSITION CORE DNB PENALTY SENSITIVITY STUOIES Runs ~aDNBR (,
+(a,c)
FIGVRE 1 TRANSITION PATTERN 1
+(a,c)
Key: ENC - ENC 15xl5 Fuel Assembly OFA 15x15 OFA
F IGURE 2 'TRANSITION'ATTERN 2
+(a,c)
Key: ENC - ENC 15x15'Fuel Assembly OFA - W 15xl5 OFA
FIGURE 3 REPRESENTATIVE AXIAL POWER DISTRIBUTION
+(a,c)
F IGURE 4 REPRESENTATIVE AXIAL POMER DISTRIBUTION
+(a,c)
FIGURE 5 REPRESENTATIVE AXIAL POWER DISTRIBUTION
+(a,c)
FIGURE 6 REPRESENTATIVE AXIAL POWER DISTRIBUTION
+(a,c)
I IU III lilt III lll IllllllllllllilllllllI lilllllllllllllIIII lilt IIIIIIII llll II llll llll IIII IIIII IIIIIIIII lllll t i ill I natl II III I lll ll IIIII II I II IIIIIIII IIIII[ll Illllll IIIIIIIIII IIII I I ! III II IIII
l0 IO .i IN('ll I X lu It(CIIFS HE EL 6 ESSM CO. NAOS III ll 5 4 46 132'IGURE 8 LOCAL QUALITY VS. ELEVATION
~ll! I ~ ~
lo 1O!5 INCII 1 x m Inn>a'S fL A ESSER CO. suocwusa. 46 132'IGURE 9 HRB-I PUBS VS. ELEVATION ill
Attachment D to AEP:NRC:0745H Response to Question on Fuel Seismic Analysis
STRUCTURAL ANALYSIS OF MIXED OFA/EXXON REACTOR CORE DURING LOCA EVENTS AHD DURING SEISMIC EVENTS An accident analysis was performed to establish the structural adequacy of the optimized fuel assembly design for use in the D.C. Cook Unit I (AEP) Plant. The specific objective was to determine the maximum fuel assembly response during a seismic or LOCA accident and to verify tnat the Westinghouse ISx15 oPtimized fuel assemblies remain eoolable. An analysis described in Section 3.5 of the Donald C. Cook Unit I July, '.1982 updated FSAR, dealt wi.th,the adequacy of the Exxon fuel assemblies when mixed with Westinghouse 15x15 standard (Inconel grid) fuel assemblies. This analysis bounds a mixed core consisting of Exxon and Westinghouse 15x15 optimized fuel assemblies. Since this plant currently conta'ins Exxon fuel assemblies, the reactor core was modeled using various core loading patterns that contained both Westinghouse and Exxon fuel assemblies. In order to perform the reactor core structural analysis, the mechanical properties of the Westinghouse 15x15 (Inconel grid type) fuel assembly design were used to simulate the Exxon fuel properties in the modeling of the Exxon fuel. It has been confirmed that this modeling of the Exxon fuel is conservative in predicting the response of the Westinghouse 15x15 optimized fuel. LOCA ANALYSIS The fuel assembly response resulting from the most limiting main coolant pipe break was analyzed using time history numerical techniques. The vessel motion for the LOCA accident produces substantial lateral loads on the reactor core, and therefore, a finite element model similar to the seismic model described in Ref. (I) was used to determine the fuel assembly deflections and grid impact forces. The reactor core finite element model which simulates the fuel assembly inter-action during lateral excitation consists of fuel assemblies arranged in a planar array with inter-assembly gaps. For the D.C. Cook Unit I Plant, 012 IP:5
fifteen (15) fuel assemblies which correspond to the maximum number of assemblies across the core diameter were used in the model. The fuel assemblies are schematically represented by individual beam .elements as shown in Figure l. A spring and lumped mass system model consistent with the model described in Ref. (2) was used to represent the simplified fuel assembly
.elements in Figure 1. The discrete masses and spring rates were calculated directly from the fuel assembly frequencies and corresponding mode shapes.
This type of model was adopted because it inherently provides an accurate representation of the fuel assembly higher natural frequencies and mode shapes. Because of the mixed core consideration; four (4) fuel assembly reactor core patterns were selected for analysis. The Westinghouse/Exxon fuel assembly relative locations for the various patterns are shown in Figure 2. These core reload patterns are consistent with typical reload configurations. I'eactor The time history motion for the upper and lower core plates and the barrel at the upper core. plate elevation are simultaneously applied to the simulated reactor core model as iIIustrated in Figure 1. The three time history motions were obtained fr om a time history analysis involving a finite element model of the reactor vessel and internals. The fuel assembly response, namely the displacements and grid impact forces, was obtained with the reactor core model by using the core plate and barrel motions that result, from a reactor vessel inlet nozzle break. The reactor vessel inlet break has been shown to produce the limiting structural loads for the fuel assembly. The maximum gr id impact forces for both the LOCA and seismic accidents occur at the peripheral fuel assembly locations adjacent to the baffle wall. The grid impact forces are appreciably lower for fuel assembly locations inward from the peripheral fuel. For the lateral blowdown case, only a small (outer) portion of the core experiences large grid impact forces. 0121P: 6
The grid maximum impact forces and fuel'ssembly maximum deflection obtained from the nozzle inlet break for the four reload patterns are given in Table I. An examination of Table I shows 'only minor differences in the fuel assembly maximum deflection and grid impact forces for the various reload patterns. Table I LOCA INDUCED FUEL ASSEMBLY fORCES AND DEFLECTION Case I* Case 2 Case 3 Case 4 Grid Wax. Impact Force (5 of Allowable Limit) 58 56 59 Fuel Assembly Max. Deflection (in) .73 .75 .75 .75 0
*Refer to Figur e 2 for co< e pattern Seismic Analysis A seismic analysis of the reactor internals was performed using a synthesized time history wave which produced a response spectra that enveloped the D.C.
Cook Unit I Plant design requirement. The time history results obtained from that analysis were used as input to the model shown in Figure I to obtain the reactor core seismic response. Since the reactor core responses obtained from the LOCA analysis were essentially the same, only three of the four core reload patterns were analyzed for the seismic accident. The grid maximum impact forces and fuel assembly maximum deflections obtained from the seismic analysis of the three r'eload reference patterns are given in Table 2. The results of the analysis show that Case I is the most limiting pattern based on the grid impact forces. The homogeneous core consisting of aII Westinghouse optimized fuel assemblies, which is Case 4, exhibited the most margin. 0121P: 6
Table 2 SEISMIC INDUCED FUEL ASSEMBLY FORCES AND DEFLECTIONS C'ase 1 Case 3 Case 4 Grid Max. Impact Force (5 of Allowable Limit) 80 75 54 Fuel Assembly Max. Deflection (in) .71 .86 .89 FUEL ASSEMBLY COMPDNENT STRESSES The stresses induced in the various fuel assembly components were assessed based on the most limiting seismic and LOCA accident conditions. The fuel assembly axial forces result'ing from the LOCA accident were the primary source of the stresses in the thimble guide tube and fuel assembly nozzles. As a result of faulted condition transient loading, the induced stresses in a fuel rod are generally very low. They were caused by bending due to the fuel assembly deflections during the seismic accident. A sumnary of the LOCA induced stresses, expressed in terms of a percentage of the allowable stress
'limits, for the fuel assembly major components is given in Table 3.
TABLE 3 FUEL ASSEMBLY COMPONENT STRESSES FOR LOCA ACCIDENT (Percent of A'llowable) f Uni orm S tresses Combined Stresses
~Com onent (Membrane/Direct) (Membrane + Bendin )
Thimble 45.5 53.2 Fuel Rod* 13.8 13.9 Top Nozzle P late 1.0 8.3 Bottom Nozzle Plate 1.0 31.6 ~Includes primary operating stress 0121P:6
The fuel assembly component stresses which result from the vertical effects of the LOCA accident were directly combined with the seismic induced stresses and a summary of the combined stresses is given in Table 4. Table 4 FUEL ASSEMBLY COMPONENT STRESS FOR COMBINED SEISMIC/LOCA ACCIDENT (Percent of Allowable) Uniform Stresses Combined Stresses Component (Membrane/D irect) (Membrane + Bendin ) Thimble 46.3 57.9 Fuel Rod* 13.9 14.6 Top Nozzle P late 1.0 8.3 Bottom Nozzle Plate 1.0 31.6
*Includes primary operating stress CONCLUSIONS Based on the grid impact forces and fuel assembly component stress margins, it is concluded that the Westinghouse 15x15 optimized fuel assemblies will remain eoolable for seismic or LOCA accidents.
REFERENCES
1. WCAP 8236, "Safety Analysis of 17x17 Fuel Assembly for Combined Seismic and Loss of Coolant Accident", L.T. Gesinski.

2. WCAP 9401, "Verification Testing and Analysis of the 17x17 Optimized Fuel Assembly", M.D. Beaumont, et.al.
0121P:6
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FIGURE t SCHENATIC REPRESENTAl'ION OF REACTOR CORE FINITE ELEMENT MODEL
00 e ~I I IQ a l SHQHHlRRIRQIRHHRIR8 l ggggggggggggggg ,HllR8%898%888%5RHl' g ggg g g gg g gg gg g
Attachment E to AEP:NRC:0745H Response to Questions on NCAP-9500 Methods and Design Criteria
Q.l Referring to Sections 4.2.1.1 to 4.2.1.3 of the NRC SER'on WCAP-9500, confirm that the design acceptance criteria stated in these SER sections are satisfied for the Cook Unit 1 15xl5 optimized fuel.
~Res onse Thhe design acceptance criteria are satisfied except for Section 4.2. 1.3(d) of the SER. The Cook Unit 1 licensing bases does not require the combining of seismic and LOCA forces during a safe shutdown earthq'uake event. Additional information on seismicILOCA is provided in the response to Question 3, Part (c).
Q.2 Referring to Sections 4.2.3. 1 to 4.2.3.3 of the NRC SER on WCAP-9500, confirm that approved methods were used for the Cook Unit 1 optimized fuel. Note and justify changes to approved methods given in MCAP-9500 which were used for the Cook Unit 1 fuel. ~Res onse Approved methods noted in the SER sections are used for the Cook Unit 1 fuel with the following clarifications or exceptions:
1. With respect to Section 4.2.3.2(d), the revised PAO fuel thermal safety model (WCAP-8720, Addendum 2) has been used in the safety analyses of all non-LOCA transients. In a July 29, 1983 letter to the NRC the following is stated, "In all cases, it was determined that the use of the fuel temperatures predicted by the revised PAD model has a slight impact on the non-LOCA safety analyses and the appropriate design bases are still met."

2. With respect to Sections 4.2.3.2(f) and 4.2.3.3(c), cladding rupture, cladding ballooning and flow blockage during LOCA incidents are accommodated in the Cook Unit 1 analyses by use of the approved 1981 large break ECCS evaluation model (WCAP-9220-P-A). Approval of the 1981 ECCS model represents a generic resolution of the clad swelling and ballooning open items stated in the WCAP-9500 SER.

3. With respect to Section 4.2.3.3(d), the seismic and LOCA forces were not combined since this was not required by the Cook Unit 1 licensing basest For additional information see the response to Q3, part (c).
Q.3 Per the NRC SER cover letter on WCAP-9500, provide the following plant specific information: a) How was the rod bow penalty accounted for? b) Confirm that the predicted clad collapse time exceeds the expected lifetime of the fuel. c) Confirm that the appropriate seismic and LOCA forces on the fuel assemblies are within acceptable bounds. d) What are the fuel surveillance plans? Res onse 3 a The Westinghouse 15x15 OFA is assumed to have identical gap closure as the Westinghouse 15xl5 LOPAR, since the parameters (fuel rod diameter, clad thickness, and grid spacing) used in analytically determining gap closure are identical. Thus, the NRC-approved full flow rod bow penalty (1) applied to Westinghouse 15x15 LOPAR is applicable to Westinghouse 15x15 OFA. This rod bow penalty is 12.5% DNBR. Sufficient margin between the safety analysis limit DNBR and the, design limit DNBR is maintained to accommodate this penalty as
J well as the transition core DNBR penalty. The additional penalty of 2.4% DNBR at loss-of-flow conditions is covered explicitly in the loss-of-flow analysis for Westinghouse 15x15 OFA. Reference to 3(a) Response:
1. Stolz, J. F., NRC Letter to T. M. Anderson, Westinghouse, "Staff Review of WCAP-8691," April 5, 1979.
Res onse 3 b Clad flattening (collapse) calculations, performed using the NRC approved clad flattening model (WCAP-8377), confirm that clad flattening will not occur during the expected lifetime of the fuel. Predicted clad flattening time for D.C. Cook Unit 1 Region 10 fuel is in excess of 40000 EFPH. Res onse 3 c Consistent with Section 4.2.3.3(d) of the SER for WCAP-9500 a plant specific analysis of several cases covering the D. C. Cook Unit 1 mixed-core configurations was done. The analysis demonstrates that grid impact forces on the OFA for the most limiting case'are:
<60% of allowable load for LOCA and <80% of allowable load for seismic However, since the D. C. Cook Unit 1 original design ba'si s does not combine gr id impact forces due to response to the LOCA and seismic events, we have not combined grid impact forces in this analysis.
Analyses also show that the major fuel assembly component stresses are less than the allowable. Hence, it is concluded the Westinghouse 15x15 OFA will remain in a eoolable configuration under the postulated LOCA or seismic events.
Res onse 3 d A routine fuel inspection program will be implemented on the irradiated and discharged optimized fuel from the initial reload region. The program will involve visual examinations on .a representative sample of assemblies from the initial fuel region at each refueling until this fuel is discharged. Visual observations will include, but not be limited to, crud buildup, rod bowing, grid strap conditions and missing components. Additional fuel inspections would be performed depending on the results of operational monitoring, including coolant activity, and the visual fuel inspections.
Attachment F to AEP:NRC:0745H Revised Peaking Factor Limit Report}}

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