ML20065L069

Forwards Supplemental Info Re Control Room Simulator Initial Certification Rept

ML20065L069
Person / Time
Site:	Comanche Peak
Issue date:	11/29/1990
From:	William Cahill, Woodlan D TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TXX-901042, NUDOCS 9012050192
Download: ML20065L069 (8)

Text

-

b.-

- s s.

MM Log # TXX 901042

+

File # 10305 W

C C

913.2 clo.

Ref. # 10CFR55.45(b) 7t/ ELECTRIC r

Wimm J. Cahm, Jr.

November 29, 1990 Larcutive Vke hesinknt p

U. S. Nuclear Regulatory Commission Attn: Director cf' Nuclear Reactor Regulation Washington D.C.

20555

SUBJECT:

. COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NO. 50-445 SUPPLEMENTAL INFORMATION REGARDING THE CPSES

.i SIMULATOR CERTIFICATION REPORT

.Ref

-TV Electric letter from W. J. Cahill, Jr. to NRC logged

[

.TXX-90291', dated August 13, 1990.

.Gentiemen:

1 With-the referenced letter, 'TU Electric. submitted-the CPSES Unit-1 Control Room

.Sirulator: Initial Certification Report.. Enclosed is add'tional information aimed at clarifying information in the initial report.

.The' attachment provides the details for placing this material in the-initial L

' report.

If you have any questions, contact Mr. Jerry Walker at (817) 897-5367.

Li

\\

Sincerely.

q William *. Cahill,Jr.

-D. R. Woodlan J

cDocket Licensing Manager s

g u'

G1.B/gj

! Attachment Enclosure q

///

c. 4Mr. R. D. Hartin, Region IV (w/o encl.)

i g

ru Resident' Inspectors, CPSES (3).(w/o encl.)

Mr.

J.~,W.

Clifford, NRR (w/o encl.)

gd

,I

. Mr.J.

Pellet. Region IV

/

j

. Mr. N.'K.'Hunemuller (NRC-ONRR)

$$f l

lc 9012050192 901!29 l

e I E

PDR ADOCK 03000445 i

F PDC q

. ju '

C y J '" ^

400 North O'ive Street LB. El Dallas, Tc.us 75201 r

.y

Attachment to TXX-901042 Page 1 of 1 The following pages should be placed in the CPSES Unit 1 Control Room Simulator Certification Report submitted on August 13, 1990:

1): r,3 move pages 6 3, 6 4, and 6 5: replace with pages 6 3, 6 4, 6-5, 6 6 and 67 2)

Insert the page titled "CPSES Annual Normal Operations Test List" in Section 8 "Four Year Test Schedule."

(

\\

[

L L'E t

isl 3

\\

u y.:

( i;I ~:'z i

l\\..

j 1

q

4 TV Electric

TURBINE STARTUP AND GENERATOR SYNCHRONIZATION This test was conducted in conjunction with the Nuclear Startup from Hot Standby to Rated

Power, This test was deemed satisfactory.

REACTOR TRIP FOLLOWED BY RECOVERY TO RATED POWER This test was conducted in compliance with ANSI /ANS 3.5-1985 section 3.1. It was a test to Y

validate the simulator perfomiance from an initial condition of - Mode 1, Reactor Power 100%,

. RCS Temperature 589 'F, RCS Pressure 2235 psig and Pressurizer level 60% - to a condition of -

t m*.

Mode 1, Reactor Power 100%. 9.CS Temperature 589 F, RCS Pressure 2235 psig cud hessurizer level 60 % y utili.ag current controlled plant Integrated Plant Operating Procedures,

. System Operating Procedur,s and Licensed Operator Training Program selected Operations Testing Procedures (suiveillance tests).

This test was deemed satisfactory.

~ OPERATIONS AT HOT STANDBY

'This test was conducted in conjunction with the Nuclear Startup From Hot Standby To Rated o

Power, Reactor Trip Followed By Recovery To Rated Power and the Plant Shutdown From Rated

Power To Hot Standby And Cooldown To Cold Shutdown tests.

i This test ivas deemal satisfactory.

7

= LOAD CHANGES -

6 LThis test was coiiducted in conjunction with the Nuclear Startup from Hot Standby to Rated

- Power. -

eThis test was deemed satisfactory.-

4 PLANT SHUTDOWN FROM RATED POWER TO HOT STANDBY AND COOLDOWN TO

.t 1 COLD SHUTDOWN c

+

T This test was ccnducted in' compliance with ANSI /ANS-3.5-1985, section 3.1.1. It was a test to 3

. validate the simulator perfomiance from an initial condition of Mode 1 (RCS temperature 589 F. RCS mssure 2235 psig and Pressurizer Level 60%) operation to Mode 5 (RCS temperature 5 200 *F, l

Keff <.99 and Power 0%) with RCS pressure < 400 psig and on' Residual Heat Removal system cooling.;

.w p:,

a f ',

This test was. deemed f atisfactory.

w r

ST4RTUP, SHUTDOWN AND POWER OPERATIONS WITH LESS THAN FULL REACTOR COOLANT FLOW This test was not conducted due to Uriit i Operating License /Teclutical Specifications not allpwing w* >

this configuration.

1 N'

-Page 6-3

'u,,

'1 i

43 e ' Jlpt-i /R 1

q ; ;, ]} WIf

TV Electric

' - CORE PERFORMANCE TESTING:

Performance of Plant Heat Balance and Shutdown Margin Calculations.

- These tests were conducted in conjunction with the Nuclear Startup from Hot Standby to Rated Power.

These tests were deemed satisfactory.

' REACTIVITY COEFFICIENT AND CONTROL ROD WOR 111 MEASUREMENTS Tt w tests are conducted in the plant using temporarily installed instrumentation therefore, they ht ot been performed on the simulator for this certification. However, similar plant startup test., nave been perfomed on the simulator by licensed operator training prior to actual plant perfonnance.

OPERATOR CONDUCTED SURVEILLANCE TESTING

.. Licensed Operator Training selected tests were conducted in conjunction with the Nuclear Startup

. from Hot Standby to Rated Power.

Rese tests were deemed satisfactory..

4 u,.'

e

.%i s

i N

\\

i f

l. I 0

i i

h7' I

+

w n',

1. [

lJ y,

lf...

\\

ll J

l

+

l Page 6-4

T1) Electric

~

TRANSIENT PERFORMANCE TESTS

- The ANSI /ANS 3.5 Appendix B2.2 transient tests have been comaleted. Attached is a copy of the completed Manual Reactor Trip test procedure and copies of test a astracts from the following -

perfonnance tests-Simultaneous Trip of All Feedwater Pumps Simultaneous Closure of All Main Steam Isolation Valves Simultaneous Trip of All Reactor Coolant Pumps Trip of Any Single Reactor Coolant Pump Main Turbine Trip (Maximum Power Level which does not result in immediate Reactor Trip)

Maximum Rate Power Ramp (100% down to approximately 75% and back up to 100%)

i Maximum Size Reactor Coolant System Rupture Comt,ined with Loss of All Off Site Power Maximum Size Unisolable Main Steam Line Rupture Slow Pr: mary System Depressurization to Saturated Condition using Pressurizer Relief or Safety" alve Stuck Open l'

r

'\\

a-

\\.

Y

-i e

g4 4

I

]

m i

,j l

Page 6 3 f

L_

.T g

Tt' Electnc MALFUNCTION TESTS 1

The malfunctions listed in the attached " Malfunction Causes and Effect" document have been tested in compliance with ANSI /ANS 3.51985, section 3.1.2. The initial condition for introducing the malfunction was as listed in the " Malfunctions Causes and Effect" document and maximum severities were used unless the minimum severity would produce the least desirable effect upon reference unit simulation. Malfunctions tested include:

(1) Loss of coolant:

(a) significant PWR steam generator leaks; L

(b) inside and outside primary containment; L

(c) large and small reactor coolant breaks including demonstration of saturation condition; (d) failure of safety and relief valves; (2) Loss of instrument air to the extent that the whole system or individual headers lost pressure and affected the plant's static or dynamic perfonnance; (3) Loss or degraded electrical power to the station, including losa of offsi'a. power, loss of emergency power, loss of emergency generators, loss of power to the plant's electrical distribution buses and loss of power to th: individual instrumentation buses (AC as well as DC) that provide power to control room indication or plant coritrol functions affecting

'the plant's response; (4) Loss of forced core coolant flow due to single or multiple pump failure; (5) Loss of condenser vacuum including loss of condenser level control; (6) Loss of service water or cooling to individual components; (7) Loss of shutdown cooling;-

Ec (8) Loss of component cooling system or cooling to individual components;

')

. Loss of normal feedwater or normal feedwater system failure; (10) Loss of all feedwater (normal and emergency);

(11) Loss of protective system channel;.

(12) Control rod failure including stuck rods, uncoupled rods, drifting rods, rod drops, and misaligned rods; (13) Inability to drive control rods;

(14) Fuel cladding failure resulting in high act..vity in reactor coolant or off gas and the associated high radiation almts; (15) Turbine trip; 016) Generator trio; W

bge 6 6

i

}

TV Electric l

I e

(17) Failuie in automatic control system (s) that affect reactivity and core heat reinoval; j

t

('3). Failure of reactor coolant pressure and volume control systems; l

(19)' Reactor trip; i

(20) Main steam line as well as main feed line break (both inside and outside contamment);

I e

(21) Nuclear instrumentation failure (s);

f I

(22) Process instrumentation, alarms, anu control system failures; I

(23) Fassive malfunctions in systems, such as engineered safety features, emergency.

)

o-feedwater systems;

-(24) Failure of the automatic reactor trip system.

3 j

LThese tests were deemed satisfactory. Minor discrepancies identified during resting have been j

scheduled for completion as shown in Section 7, 1

J

.i

A " Malfunction Cause and Effects" listing for all cenified malfunctions is attached.'

j i

n 4

i 1

i i?

a

'l

?s L

j q

1 1

[.

jj

.)

.{

l 1

h.

..j f

4 i

I l

Page 6-7 u::

1 4

CPSES ANNUAL NORMAL OPERATIONS TEST LIST Plant Startup - Cold Shutdown To Hot Standby Nuclear Startup From Hot Standby To Rated Power Turbine Stanup And Generator Synchronization

. - Reactor Trip Fouowed By Recovery To Rated Power Operations At Hot Standby Load Changes Phuit Shutdown From Rated Power To Hot Standby And Cooldown To Cold Shutdown Core Perfonnance Operator Conducted Surveillance Testing On Safcq Related Equipment Or Systems

,.2 q,

j-p

, t.... -..

i

.ji

'l c,.

5 i

4

.......................