ML20246P597
| ML20246P597 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/22/1989 |
| From: | William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-89139, NUDOCS 8903280261 | |
| Download: ML20246P597 (55) | |
Text
_ _ _ _ _ _ _ _
sr
's EliMRMIEW M F ~-- :
Log # TXX-89139 c
Fi1e #
10010 r
x 914.2 Ref. #
10CFR50.34(b) 7UELECTRlC March 22, 1989 W. J. Cahm f.secuane %ce l'residernt U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET N05. 50-445 and 50-446 INITIAL STARTUP TESTING PROGRAM Gentlemen:
TV Electric has completed an extensive review of the CPSES Initial Startup Testing program.
The objective of this review was to ensure that the CPSES has an efficient and effective prcgram that tests all required systems, provides the requisite data to confirm the design of the systems, and provides the necessary experience for safe plant operations.
The following information was included in the review:
1.
Information supplied by Westinghouse and consultants experienced in startup testing.
Since 1968, Westinghouse has been involved with the design, construction, startup, and operation of twenty-six domestic four loop plants.
Twenty of these plants have started up since Regulatory Guide 1.68 Revision 2 was issued.
2.
NRC and industry testing experience (e.g., LERs, INPO reports, and HUREG 1275, " Operating Experience Feedback Report - New Plants")
3.
First hand observations by various experienced TV Electric personnel of the conduct of startup test programs at several recently licensed plants.
4.
FSAR commitments associated with startup testing and the available guidance for the performance of the testing (e.g., Regulatory Guide 1.68, ANSI /ANSstandards) 5.
Information related to equipment and test configurations that have historically been the source of testing difficulties at other facilities (such as B0P systems).
8903280261 890322 PDR ADOCK 05000445 I
A PDC 400 North Olive Street LB 81 Dallas, Texas 73201
}
TXX-89139 March 22, 1989 Page 2 of 2 Since the issuance of Regulatory Guide 1.68, a significant amodnt of plant experience has been gained from the conduct and evaluation of results of startup testing, which supports modifications for a more effective test program. TV Electric, Westinghouse, and various startup testing consultants have jointly reviewed the CPSES Initial Startup Test program and have identified several methods for improving the program.
These methods include:
1, Conducting thorough reviews of test procedures and performing dry runs of planned tests (plant simulator, desk top, and actual plant equipment when appropriate and available), especially(tests that are known to have caused difficulty at oi.her facilities e.g., feedwater control),
2.
Eliminating redundant and/or unnecessary testing, 3.
Incorporating alternate / improved test methods and acceptance criteria, 4.
Consolidating required testing to eliminate or reduce unnecessary plant transients, and, 5.
Revising the power levels and test sequences to minimize the potential for reactor trips and unnecessary actuation of ESF systems.
l In order to implement these impro.ements some modifications are required to the FSAR. Attachment 1 provides marked-up FSAR pages and Attachment 2 provides the technical justifications for the planned revisions to initial Startup tests covered by FSAR Table 14.2-3 as well as changes to some related l
FSAR sections.
These revisions will be included in an upcoming FSAR amendment.
In addition, the startup test organization has improved several areas of the testing program to anticipate and prevent or mitigate testing problems should they arise. Consideration is also being given to establishing an appropriate power plateau holdpoint to assess test results, test procedures, and equipment and personnel performance, prior to proceeding to higher power levels.
As recommended in NUREG-1275, TV Electric is striving for "... a deliberate, evenly paced, thorough and well planned... startup test program..."
Sincerely, 1
/"
William J. Ca 11, Jr.
I BSD/bsd Attachments c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) w__-_____-____-_-___----_
Attachment I to TXX-89139 0
March 22, 1989 Page 1 of 35 CPSES/FSAR E
e 3.98.2.1.2 Dynamic Transient Response Testing l57 Instrumented dynamic transient respnse testing augmented by visual l66 observation for the specification flow medes analyzed will be l
performed on the following piping systems:
l l Q112-System Transient Tested l66 1.
Main Steam System Main Turbine Trip at l66
- ) 50';
- t;;; f1:
- -d l66 l
--4)- 100% steam flow l66 l
2.
Safety Injection System Safety Injection Pump l66 Trip During Injection l66 I
3.
Main Feedwater System Main Feedwater Pump Trip l66 I
4.
Auxiliary Feedwater System Auxiliary Feedwater Pump-l (including Steam Supply Line)
Trip l66 1
I 5.
Pressurizer Relief Valve Pressurizer Relief Valve l 66 Discharge Piping of Reactor Actuation l66 Coolant System l66 I
6.
Service Water System Service Water l66 l
Pump Trip & Restart l66 j
lQ112.12 l During the dynamic transient response testing for the specified piping l 66 1
systems,thepipingwillbeinstrumentedtomeasuretransientloadsatl l
selected locations and visually observed at other selected locations l
l to ensure that severe vibrations do not exist.
The instrument data l
will be compared with the analytically predicted values.
l 3.98-17 Seemem maao
N Attachment l'to'TXX-89139 March 22, 1989 CPSES/FSAR Page 2 of 35 Table 14.2-3 (Sheet 3)
REACTOR C00LAf1T SYSTEM Fl.0W COASTDOWN TEST TEST
SUMMARY
CPSES OBJECTIVE
'l52 To measure the reactor coolant system flow rate decrease subsequent to l 52 a simultaneous trip of all. four reactor coolant pumps, and to measure l
the delay' times associated with assumptions of.the loss of flow l
. accident analysis.
l l
l PREREQUISITES l 52 1)
The reactor is in the hot standby condition.
l52
- 2) Applicable portions of the Reactor Coolant System Flow Te<' have l52 l
been satisfactorily completed, and any pressure damping r,. ices, l
I (
if installed for that test, have been removed.
l TEST METHOD l52
- 1) With the reactor coolant system in the hot standby condition.
l52 simultaneously trip all four reactor coolant pumps.and measure the. l rate at which reactor coolant flow decreases.
l
- 2) Determine the delay times associated with the low flow reactor l52
.)
trip circuitry.
l 1
i ACCEPTAtlCE CRITERIA l52 The rate of reactor coolant flgw decrease, upon tri,pping of all four l52 reactorcolantpumps,i[.,..I.^.
+h
~, _ '
+ ' _. '. ' '. ^..d^
^
-a m ~ed in the loss of flow analysis of the FSAR, and the associated l
delay times are within acceptable limits.
l t
?
November 20, 1987
~
, to TXX-89139 CPSES/FSAR
~ March 22, 1989-Page 3 of 35 Table 14.2-3 (Sheet 12)
AREA RADI AT10fl f10filTORillG Atl0 RADI AT'I0ft SURVEYS I
- TEST SUIit1ARY CPSES i
OBJECTIVE l6 l 0423.16
- To veri fy thc perfor;;;aric;. of th; ar ca, adiat ica.::an ita: i :g equip::=t l' 6 e 4 radiation shielding effectiveness by measuring radiation dose-l levels at preselected locations within the plant during low, l
intermediate and high reactor power level operation.
-l P_RERE0VISITES 1.
The radiation survey instruments are calibrated against known-sources.
l0423.16 7
is m
-,a:,..s
.m i.,
p_
' t cd 4,s t Liwan j 6-
. a :'
c 2,/.
Reactor power level is established as. necessary.
3 TEST !!ETHOD i
l_Q423.16 1.
At specified steady state power levels between zero and 100
'l6 percent power, measure radiation levels at preselected locations l
within the plant to determine effectiveness of the-r-adiation '
l shielding.and crea radict'_
m;.. i t o r s.
l 2.
Upon completion of the radiation surveys, check the calibration 'of i
the survey instruments.
ACCEPTAllCE CRITERIA lQ423.16
(
The measured radiation levels are within limits for the zone
{6 designation of each area surveyed.
Th
,,,, uJ i a t i c;;
m.; i t t,,;;g l
eqm p --+ p q ~ !y *~u - t -
+i,-
A 4, + 4 ~, 1;
,-s pa - ~
l
"~d mc c itb d^:ign ;p ;ifi; t ur3 l
November.20, 1987
', to TXX-89139 March 22, 1989 Page 4 of 35 CPSES/FSAR Table 14.2-3 (Sheet 15)
CONTRUL ROD REACTIVITY WORTHS TEST
SUMMARY
CPSES OBJECTIVE ve.ri f y +h e. cle s i g n lQ423.lf, ro Jo-det-ermine the differcrtia! and integrai rod worths of the control l6 and shutdown RCCA banks,and the -ir mum berer concentratier required l
i for,na i n t a i a4mp44te-reac t o r s h u t dc.in.s i t h t he mo s t r c a c t i v e ",CC A s t u c k l 4n the 'u'1 out positienm l
PRERE0VISITES 1.
The reactor is critical, in the hot, zero power condition.
TEST METHOD
~
l Q423.16 1.
With the use of boron concentration sampling data, rod p+osition l52 ye ei F 'a hf-indication and reactivity measurement, dctcrmine the u,ificitolioi l
worths of the control and shutdown RCCA bankst U+iliyah
+hc.
l munod o A Bad heheac-(or hun 0,.,, % % n t=cua.yc)- l Q423.16 2.
Calculate t4 c integral.orths Of the cOatici 6iid sliuluvwn RCTA l6 banks.
J
-b-W44h--the-most---re ac t4v e -R CC A-i n-t he - f u ll-ou t-pos it ion, de t e rmine-l52 t-he minimum beren concentntier required te maiat? ia the reacter l
shutdown.
l ACCEPTANCE CRITERIA The rod worths are determined to be within design specifications, and l 52
- he terni in+egr=1 reac+ M ty -~+ h of =11 cea+rn1 s a a c h " + a -- u-t s l
4ew-the h i g hes t--wor-t4: Stuc! 9CCE is greeter th'n er equ2 t o-t hw l
1 hluc ced in--the-Safety ^"alysi:.-
l November 20, 1987
q
, Attachment I to TXX-89139 March 22, 1989 Page 5 of 35 CPSES/FSAR l
Table 14.2-3 (Sheet 19)
R0D DROP TESTS
{
TEST
SUMMARY
CPSES OBJECTIVE To determine the rod drop time of each full-length rod cluster control l 52 assembly (RCCA) under cold--nc #!cc., <c!d f u l ' # !c<c.nd hot full flow l
Reactor Coolant System conditions.
l PREREQUISITES 1.
Plant conditions are established as required.
2.
Rod position indication is functional.
TEST METHOD
(
j0423.16 1.
Withdraw each full-length rod cluster control assembly, interrupt l6 the electrical power to the associated tod drive mechanism, and l
measure and record the rod drop time.
This test is performed l 46 with the reactor cold "m
r i m ]g,,,
r"11
'1~
and hot full flow l
conditions.
l
{
af lea 3+ are e.
l l0423.16 Perform W additional rod drop tests for each rod whose measured j
2.
l52
' drop time deviates from the mean for all rods by more than two l
standard deviations.
l ACCEPTANCE CRITERIA l
The rod drop times are acceptable in accordance with plant Technical l52 Specifications.
l l
I November 20, 1987
~
Attachment I to TXX-89139 i
March 22, 1989 CPSES/FSAR Page 6 of 35-Table 14.2-3 (Sheet 20)
FLUX DISTRIBUTION MEASUREMENTS TEST
SUMMARY
CPSES OBJECTIVE To determine the reactor core power distribution $.for cricus cor, trol red configuraticas.
)
l PREREQUISITES 1.
Incore instrumentation and proces. computer are operable for incore flux mapping.
I 2.
Reactor is critical and power level is established as necessary.
]
TEST METHOD we AH R ods 004 (Aco) j an Complete incore flux mapf for eeeh of the following control rod l 66 3
configuration ( with reactor power stablized et approx %:tely 3 pcrcer,t be,lo w te 5 percent:
l Q121 All Rods Out (AR0) l6 l
C-ent+ol-Bank-D-inserted l6
]
Control Banks D, C, 9 at the het :ere pc'c:cr
]6 insertier limit l
~
l66 J
def'"' d C*' A 's
- e* w ernc ~ * '
N v1<. ARO
's cov4rol S ou n k D a bove.
in o skes w Adra w~
nnel a ll e f < < b. k s nf
- 2. 2 e sk ps.
~Amendmenf 66
. January 15, 1988
w
- to TXX-89139 March 22, 1989 CPSES/FSAR Page 7 of 35 (Sheet 21)
ACCEPTAllCE CRITEPIA Thecorefluxdistributionsindicatedbythefluxmapjareacceptable l52 in accordance with plant Technical Specifications where applicable.
l 1
November 20, 1987
,e' Attachment l'to TXX-89139 March 22, 1989
{
Page 8 of 35 CPSES/FSAR i
Table 14.2-3 (Sheet 22)-
CORE PERFORMANCE EVALUATION-TEST
SUMMARY
CPSES-OBJECTIVE l.Q423.16 To verify the operating characteristics of the core and the
[6 calibration of the flux and temperature instrumentation during power l
escalation.
l l-PREREQUISITES-1.
Reactor power level is established as necessary.
a TEST METHOD l
1.
At steady state power levels of 30, 30, 75, 90 and 100 percent, record Reactor Coolant System paraneters,and Incore data f o EIux maps et " '#*
A4 50 75 ' a ard
'o o 3
eeretaf. l Q423.16 2.
Analyze the data obtained at each power level to determine core l6 I
performance margins and verify flux and temperature l
instrumentation calibration.
l l 52 ACCEPTANCE CRITERIA lQ423.11 lQ423.16 lQ423.8 The core performance margins are within design predictions for normal l52
' 9 Acr ' rod configurations and the calibration of the flux and l
temperature instrumentation has been verified.
l November 20,-1987
.,' Attachment 1 to TXX-89139 March 22, 1989-CPSES/FSAR page 9 of 35-Table 14.2-3 (Sheet 23)
UNIT LOAD TRAflSIENTS TEST
SUMMARY
CPSES OBJECTIVE To demonstrate satisfactory plant transient response to various specified. load changes and trips, to monitor the behavior of reactor control systems during these transients,Jand, if necessary, optimize '
the reactor control system setpoints.
PREREQUISITES 1.
Reactor power level is established as. necessary for each transient.
2.
(,
.All reactor control systems are operational end. their setpoints have been set to their recommended values.
TEST METHOD l52 1.
Initiate a step change in power level of 10 percent and monitorf Reactor Coolant System behavior in response to the transients.
This test will be performed at power levels of hpercent, 70 norennt, and 100 percent.
2.
Monitorplantresponsetoa50percentloadre'duc'5 ion,frompower
~
levels of 75 percent and 100 percent.
j 3.
Monitor plant response to lanttripfirampoeserlevelsupto100 percent.
/
4.
If necessary, adjust the reactor control syst.em setpoints until optimal response is obtained during subsequent test performance.
\\.
November 20,l1987
V Attachment I to TXX-89139 March 22, 1989 CPSES/FSAR Page 10 of 35 k.
(Sheet 24)
ACCEPTAflCE CRITERIA Plant response to the unit load transients is acceptable in accordance with design specifications, and the Reactor Control System parameters reach steady state values without appreciable overshoot or oscillation subsequent to a step change.
(
s.
November 20, 1987
m
' Attachment 1~to TXX-89139-
.o March 22, 1989 Page 11.of.35 CPSES/FSAR1 Table 14.2-3 (Sheet 25)
REMOTE SHUTDOWN TEST
SUMMARY
CPSES 0_BJECTIVE l043.16 To demonstrate the capability of performing a safe plant shutdown, l6 maintain the plant in a hot standby condition, and to demonstrate the. [
ability to cooldown from hot standby to-cold shutdown conditions from l f
outside the control room, using the minimum shift crew.
Verify that.
l 52
)
the Remote Shutdown Panel selector switches properly tr'ansfer control l
from the Control Room to the Remote Shutdown panel.
l k
PREREQUISITES l
l6 l
l0423.16 I
I 1.
The equipment and instrumentation associated with the Remote-l6 l
Shutdown Panel are available for achieving and maintaining the l
f plant in a hot standby condition.
l PI 6 t 34nemfer r ocue r.
l0423.16-o A
2.
The macter is at a power level greater than 10%g ut less than l6 b
25% reado< e m en Car A t-reader 4 r m enet ho4.wan gord..., o f
- 1. 4e34, f
P * *I 'S * * ^ 3
- b It b
- f' S I* Ab l
4 5
3, For As coeldean porl'** s A i
)
TEST METHOD CA^ C/ d *1 I
y n e rdor*
1.
With the reacter phnt at greater than 10 percent power, perform a l 52 safe'shutdownoftheplantfromoutsidetheControlRoomusingthel minimum shift crew.
l 2.
Check functioning of instrumentation, controls, interlocks and i
a1 arms. C<<M ma y he h& Co < ere ce/ pre up % c:1 onsI i n +.s.
j l 0423.16 ~
3.
Demonstrate the capability to achieve and maintain the plant in a j6.
hot standby condition from the Remote shutdown panel for a minimum l of 30 minutes.-
l Novembe r_20_,;_1198_7___
i Attachment I to TXX-89139 March 22, 1989 Page 12 of.35 CPSES/FSAR Table 14.2-3 (Sheet 26) l Q423.16 l
.4.
Demonstrate the potential for cooldown to cold shutdown conditions l}6 j
by placing the residual heat removal system into service and l
reducing the reactor coolant temperature to approximately l.
3000F.
l I
ACCEPTANCE CRITERIA l0423.11 l-Q423.16 Transfer of control to outside the Control Room can be. achieved in l6 4
accordance with design requirements, remote shutdown ins.trumentation, l
controls, alarms and interlocks function properly.
The. potential l.
ability to perform a safe shutdown, to achieve and maintain hot l
standby conditions from~outside the Control Room has been l
demonstrated.
The potential ability to cooldown to cold shutdown.
l52 conditions from outside the control room has been demonstrated.
l 2
1 I
i November 20, 1987-a
- to TXX-89139 March 22, 1989 CPSES/FSAR Page 13 of 35 Table 14.2-3 (Sheet 33)
AUTOMATIC REACTOR CONTROL SYSTEM TEST
SUMMARY
CPSES OBJECTIVE l~52 Demonstrate the ability of the Automatic Reactor Coolant System.to l-52.
return' reactor coolant temperature to the programmed setpoint and to l
maintain'it.
l_
l-PREREQUISITES l41
.gg,4 l04232' j
l'.
The reactor is at approximately C; power (757 fer the.second part-l 41
-ef the test).
j lQ423.2 2.
Reactor Coolant System temperature and pressure'are stable.
.]41 l Q423.2
-(
- 3. ' Pressurizer pressure and-level control are in automatic.
l41 lQ423m 4.
Steam Generator' level control is in automatic.
l41 TEST METHOD 50%
.l0423.2 1.
At -90s power,, raise Reactor Coolant. System temperature and observe l 41 the Automatic Reactor Control Sy. stem response.
Then lower j'
Reactor Coolant System temperature and observe the Automatic-l Reactor Control System response.
~ ~ ~
j lQ423.2 2.
At 750 pc.act, ra i 0 900 c t er C00 ! 2 ^ t Sys t er t empentur+-and-obwr-v4 l. 41 the hte:ratic 9erter Centrol S):te'" ~cp^ase.
ne- !crer l
Per t ar Coel?.nt Syste!n te'nperat" " nnd ah"~~
+h e ^utematic l
- "coctor Ccatrel Syster espe"re l
ACCEPTAUCE CRITERIA
'N.
.lQ423.2 The Reactor Coolant System Temperature (Tavg) returns to'within l 41 1.50F of the setpoint (Tref) following the positive and negative l
temperature transient.
l.
Novamhar ?n 1807
t.
' Attachment l'to TXX-89139 March 22, 1989 Page.14 of'35 CPSES/FSAR l
Regulatory Guide-1.68 Preoperational and Initial Startup Test Programs for Water-Cooled-Power Reactors Discussion l0400.3 ThetestingactivitiesconductedaspartoftheInitialTestProgram,~l11
-l as described in Section 14.2, and startup program comply with the l
intent of this Regulatory Guide, Revision 2, dateC August 1978, with l
f
~
the following exceptions and justification.
- l 1.
Appendix ~ A subparagraph 1.c l.- 11 i
1 I
l As identified in Table 14.2-2, Sheet 42A of 60.of the FSAR, the l11 response time acceptance criteria of the various logic channels l
will be consistent with Technical Specifications requirements.
l TheReactorTripSystemResponseTimeisdefinedinthetechnicall
(
specifications as the time interval from when the monitored-
. l parameter exceeds its trip setpoint' at the channel sensor until l
loss of stationary gripper coil. voltage.
The-accident analysis l
accounts for conservative values for delay times, setpoint dri.f t, l etc.
Therefore, it is not necessary to account these delay l
times in the test methods or acceptance criteria.
. l lQ400.3 l
In the design basis of the Comanche Peak Steam Electric Station, l11 j
the effects of anticipated transients without scram (ATWS) were l
not considered as described in' FSAR section~ 4.3.-T.7,.therefore~ a l
commitment to test any design features to prevent or mitigate l
ATWS can not be made at this time.
l-l-Q423.22 i
2.
Appendix A subparagraph.l.d.1, i.e.6, l.e.g, 1.f.1, 1.j.2 &-
l11 1.j.17 l
k 1 A(B')-31 February 15, 1988
Attachment I to TXX-89139 March 22, 1989 Page 15 of 35 CPSES/FSAR l
11 1
There are no safety related functions of the turbine control and j
l bypass valves as discussed in FSAR Section 10.2.2.7.7.
The D
l circulating water system also has no safety related function as i
described in FSAR Section 10.4.5.
Therefore these systems are l
not included in the preoperational program.
However, as I
identified by Section 14.2.1 these systems are intended to be I
acceptance tested in accordance with applicable startup l
administrative procedures.
11 l
The feedwater control system is not required for safety but does l
have an interface with the protective system.
These interfaces I
are exclusively part of the feedwater isolation valves ability to I
close upon the actuation of the proper logic and is covered l
within preoperational test summary FSAR Table 14.2-2, Sheet 50.
l Section 0400.3 l
11
}
14.2.7 has been revised to delineate tests to be perfonned on the l
feedwater cont rol system.
1 11 l3.
Appendix A subparagraph 1.d.3 Q400.3 l
11 l
The accident analysis concerning the inadvertent depressurization j
of the reactor coolant system is discussed in FSAR Section l
3.9N.I.1, Upset Condition number 5.
Subsection 5.b lists the l
l condition of the inadvertent opening of one pressurizer power l
operated relief valve (PORV).
The analysis states that the i
l limiting case is the actuation of the pressurizer safety valve._
l Tliis is classified as an Upset Condition with no operational
~
{
l l
impairment.
The design parameters listed in Table 5.4-16 of the l
FSAR indicate that the relieving capacity for the pressurizer m
l power operated relief valve is one-half the capacity of the l
pressurizer safety valve, 210,000 lb/hr vs 420,000 lb/hr.
l Therefore there is no int"ntion of performing capacity tests of l
the pressurizer power operated relief valves during the startup
]
phase.
February 15, 1988 1A(B)-32
Attachment I to TXX-89139-March 22, 1989 Page 16 of 35 CPSES/FSAR
. Testing of these valves is listed in Table 14.2-2, Sheet 53, of l.11
(,
the FSAR as a preoperational test.
l The accident analysis covering the opening of a steam generator
- l. 11 power operated relief valve is-similar.
The analysis is-l~
described in FSAR~Section 15.1.4.
The accident analysis uses a l
value of 964,800 lb/hr 0 1200 psia for the steam flow.
Table l
l 10.3-3 in the FSAR lists the. steam flow of the steam generator-l power operated relief valve at 420,000 lb/hr.at 1107 psia, j
Although the design pressures are different, the conservatism l
~
allowed"Sy the accident analysis parameters indicates no need to -l test the capacities of the steam generator power operated relief l
l valves during the startup program.
l
)
l The testing of the steam generator power operated relief-valves.
is presented in FSAR Table 14.2-2, Sheet 49.
4 4.
Appendix A subparagraph 1.k.2,3 The equipment identified'in the above paragraphs of Regulatory
'l 18 Guide.1.68 is' calibrated and functionally tested as part of the l
instrument calibration program for the TUGC0 Chemistry and Health ['
Physics section.
The calibrati_on and functional testing is l
performed and documented in accordance with approved station l
calibration procedures.
Therefore, TUGC0 Startup will not l
l perform additional testing, in the form of a preoperational test, l on this equipment.
l 5.
Appendix A subparagraph 1.n.11 The design of the Instrument Air System as referenced in FSAR J
Section 9.3.1.1 is not nuclear safety class because it serves no safety function required hv Appendix A of 10 CFR 50.
Regula tory Guide 1.68, "Preoperational and initial Startup Test Programs for i
' Water-Cooled Power Reactors" and Regulatory Guide 1.80, k
1A(B)-33 FebriTary 15, F988
Attachment-l'to TXX-89139 March 22, 1989 Page 17 of 35 CPSES/FSAR i
"Preoperational Testing of Instrument Air Systems" both' specifically state their scope is limited to safety-related
]
systems.
Therefore, using.the stated design.there appears to be no requirement to include the instrument air system in the preoperational testing program.
However, as identified by l
Section 14.2.1, this system is intended to.be acceptance. tested in accordance with applicable startup administrative procedures.
36 l 6.
Appendix A subparagraph 1.o.1 36 l
The vendor has perfonned applicable load testing on the head I
lifting and internals lifting devices for 125% static loads.
36 l7.
Appendix A subparagraph 4.t 54 l
tlatural circulation tests have been successfully. completed at l
McGuire Unit 1, Salem 2, Sequoyah I and other Westinghouse plants
}
similar-to CPSES.
It is unnecessary for CPSES to compare f. low l
(without pumps) and temperature data to that of these plants
,}
l since no design differences exist which would significantly l
effect natural circulation capabilities.
Typical ~ natural l
circulation characteristics for 4 loop Westinghouse plants are j
given in.WCAP-8460, " Natural Circulation Test Report for Zion l
Station Unit 1."
However, in order to verify natural
[
circulation cooldown and boron mixing capability per requirements l
of Branch Technical Position RSB 5-1, CPSES Will reference test l
results from Diablo Canyon Unit 1, or other similar Westinghouse l
plants, if test results are found to be acceptable.
I f-l unacceptable, CPSES will perform a test prior to or during-l startup following the first refueling outage.
See Appendix 5A l
for further discussion.
'h l
February 15, 1988
'1A(B)-34
~
~
'. to TXX-89139-March 22, 1989
. Page 18 of 35 CPSES/FSAR 8.
Appendix A subparagraph 5.x.k~
The most influential contributor for this transient is the value l, 41 of moderator temperature coefficient of reactivity, which has a.
.l relatively low value at beginning of core life.
Since this l
~ _'
parameter is determined in other startup tests, thus validating l
thesafetyanalysis,theperformanceofthistestprovidesnonewj' information needed to verify the plant design.
The transient l
does introduce the potential for thermal stress damage to the j
steamgeneratorfeedwaterinletnozzlesanditexpendsoneofthel analyfed thermal cycles.
Therefore, we do not intend'to perform -l a test to comply with this subparagraph.
l 9.
Appendix A, Subparagraphs 4c, Se, 5f and Si l-66 The pseudo ejected rod tests and pseudo dropped rod tests l 66 referenced by these sections of the Regulatory Guide have been l
successfully performed at plants of similar design to Comanche l
[
Peak.
Previous tests on similar facilities indicate little new.
l information is generated by the performance of these tests.
l Additionally,vendorpredictionsindicatethataviolationofthe'l F-delta-H Technical Specification may occur if the 50*r power l
-l pseudo dropped rod test is performed at Comanche Peak.
Recent l
industry experience indicates there is an increased potential for l causing severe xenon transients by performing these tests.
l Based on these reasons, we do not intend on' performing tests to l
comply with Appendix A, Subparagraphs 4c, Se, 5f and 51.
[
- 10. Appendix A, subparagraph 5.d.
l41 l
CPSES plant design does not include part-length control rods.
l41 The ability to control core xenon transients is a design feature l
of the Westinghouse Huclear Steam Supply System and has been l.
k 1A(B)-35 February 15, 198E
y Attachment I to TXX-89139-March 22, 1989-Page 19 of 35 CPSES/FSAR
^
41 l'
demonstrated in numerous operating pressurized water reactors.
l In addition, compliance with Technical Specification 3/4.2.1,
-)'
l
" Axial Flux Dif ference", ' ensures proper poner and flux l
distributions. On these bases, CPSES does not intend to perform an initial Startup Test to comply with subparagraph 5.d.
41.
l 11.. Appendix A, subparagraph 5.i.i.
41 l
The performance of this test provides no new information needed-l to verify the plant performance during design transients'. Trip l
of-the reactor coolant pumps result in a reactor trip with flow l
coastdown, as' verified in the Reactor Coolant System Flow.
j l
Coastdown Test, providing sufficient heat removal to ensure DNBR l
l does not decrease below 1.30.
Performing this test expen'ds one f
l of the analyzed transients and results in unnecessary cost and l
l down time for the utility.
l 41 l12. Appendix A, subparagraph 5.u.
j g
)
\\
41 l.
Operability and response times of the main steam isolation valves l
will be verified in hot standby (mode 3) rather than at the l
recommended 25% power level.
Testing at hot, zero power will l
result in more conservative results and will eliminate the j
l unnecessary pressure and steam flow transients which would l
otherwise be induced.
t 41 l13.
Appendix A, subparagraph 5.ci.m.
41 l
The performance of this test provides np new information needed l
to verify the plant performance during design transients.
l Closure of all Main Steam Isolation Valves from 100't power causes l
a turbine trip and a reactor trip. A turbine trip'and' reactor l
trip from 100% power will be performed during initial startup l
testing.
Closure of the MS!v's may cause the operation of tne 1
()
l l
...g emW February 15, 1988 1A(B)-36
- to TXX-89139' March 22, 1989 Page 20 of.35 CPSES/FSAR pressurizer and steam generator power operated relief valves l 41
(
and/or safety valves which may then requ re repair and l
unnecessary down time for the utility.
This' test would expend.
l one of the analyzed pressure transients for the reactor coolant l
~
system and steam generators and therefore will not be performed l
at CPSES.
l Appendix A, subparagraph 2.h O l 46 14-uo s: cow wb c.ocb rocc 94_ow y coup Hot no flowgro,d drops do not provide any additional useful data.
l46 (3y TeJnical Se.ofoddThe *$[t 4= for het ac fico conditions eic lun unse,.ative l.
N((4 Ij'ng,tfg,fico conditions.
We do not intend to perform c.e4M ao fic~
h
,w trf,3 Aet ed RcP3 hot no flow rod drops.
l b
pacawy ad M svws r*Q"f, fi%
. r *. co*
$ "',7/,,fti.a
- 15. Appendix A, subparagraph 5.h l52 t oo.I, hen s Rod drop times are measured during pre-critical testing at-cold l-52
,+
i nn flow, colri in11 f %, and hot full-flow conditions.
There is l
(
noprovisioninthedesignofCPSEStoallowfordeterminationofl rod scram times following normal plant trips.
These tests meet l
the intent of subparagraph 5.h.
l
.16. Appendix A, subparagraph 5.j
.l52 The CPSES design does not include partial scram or rod runback l52 features.
Therefore, an Initial.Startup Test will not be-l performed to comply with subparagraph 5.j.
l
- 17. Appendix A, subparagraph 5.1 l52 The CPSES residual heat removal (RHR) system does not include a l52
" steam condensing mode" of operation.
Also, the " reactor core l
l isolation cooling (PCIC) wstem" pertains to Boiling Water l
Reactors.
Since these features do not exist at CPSES, they l
k i
1A(B)-37
~ ~ February 15, 1988 j L_______________._______________________
Attachment I to TXX-89139 March 22, 1989 Page 21 of 35 CPSES/FSAR 52 l
cannot be tested during the Initial Startup Test program.
The
)
l remainder of paragraph 5.1 will be complied with.
]
65 l18. Appendix A, subparagraph 5.a 1
65 l
The power coef ficient measurement test method recommended by l
Westinghouse cannot be used because it requires installed l
instrumentation (narrow range hot leg temperature) which is not s
l part of CPSES design.
Power coef ficient measurement results, l
obtained during the startup of several Westinghouse plants, have l
l verified consistency between design calculations and measurement l
data in plants having fuel similar to the Comanche Peak design.
l CPSES's initial Startup program includes measurement of the l
critical RCS boron concentration under the conditions of all rods l
out, hot full power, equilibrium xenon.
From this, the core l
reactivity balance is verified.
This test is considered
\\
sufficiently accurate to detect any significant differences l
between the designed and as-built core.
This test is in l
addition to the minimum specifications of the Regulatory Guide.
65 l19. Appendix C, subparagraph 4.h.
l 65 l
The power ascension program for CPSES requires flux maps to be l
i l
taken at each power plateau.
The values for Fxy (heat flux hot l
II channel f actor) and F H (nuclear enthalpy rise hot channel l
factor) obtained from these flux maps will be compared to and
'l l
evaluated against Technical Specification limits.
This l
evaluation will be used in place of extrapolating DilBR and linear
~
l heat rate values.
This action is consistent with the Reference l
Startup Document supplied by the fuel vendor.
Station l
Administrative procedures will require management review and l
approval of test results prior to ascending to the next power i
plateau.
h (***'A
~
2, 0,
ses.A<o~
C.. S, A PPEu b'd so M%
pwu kalix e) odo fa c a.f, o,
<}
tnac e re ap +roh February 15, 1988 1A(B)-38
_m Attachment I to TXX-89139 March 22, 1989 Page 22 of 35
- gg g
y
'T E t: TIN C, M L L.
mbt C E.
PLRFoKmEb AT TH e' RE G.
Gu TOE s uc c cs 7py
- 2. 5 */a Qo%) Fom E tt AS % o n s '.
tzesc. wit THenna powce Theteststhatarepresentlyprescribedtobeconductedatthe30%fRTP) plateau include the following:
Core Performance Evaluation.
Unit Load Transients (10% Step Load Change).
Automatic Reactor Control System Test.
The core performance data that could be obtained at 30% RTP is utilized for gross calibration adjustments of the Nuclear Instrumentation System (NIS) prior to power escalation to 50% RTP.
This activity will be performed at 25-30% RTP as a hold prior to escalation to 50% RTP.
The flux distribution measurement at 30% RTP will not be performed unless the peaking factors measured at low power do not support escalation to 70% RTP, the NIS trip setpoint for the 50% RTP testing plateau.
This is per the direction of RG-1.68 Appendix C, paragraph 4.h.
The unit load transient at 30% RTP was intenoeo to oe a precursor to the same transient test for higher power levels, and the 50% load reje : ion transients. However, a load change transient at 30% RTP does not provide useful predictions about whether acceptable control system performance would be obtained at the higher power levels.
A step load l
decrease from 30% RTP results in the main feedwater control valves trying to operate at the lower acceptable end of their operating range, leading to poor control system response.
Making control system adjustments (especially to steam generator level control) to address the operating concerns at 30% RTP has a high probability of causing more significant control problems at the higher power levels.
This was noted during startup at vedusplants where setpoint adjustments made at the 25-35% RTP resulted in unacceptable performance during the 50% load rejection transients.
~
The Automatic Reactor Control System test is intended as a precursor to the Unit Load Transient test and is performed at 30% RTP..
It is designed to ensure that the automatic rod control system can restore the Reactor Coolant System (RCS) temperature to within a 1.5 Deg-F deadband of the reference temperature.
Prior to 50% RTP, proper operation of this function would be demonstrated by observation during the normal power escalation, where the control rods will be in automatic and already controlling the RCS temperature to within the deadband.
W W
'. to TXX-89139-March 22, 1989 P, age 23 of 35 CPSES/FSAR 1
functioning during reactor operation:
1.
All materials are procured to specifications to attain the desired standard of quality.
2.
A spider from each braze lot is proof tested by applying a 5000 pound load to the spider body, so that approximately 310 pounds is applied to each vane. This proof load provides a bending moment at the spider body approximately equivalent to 1.4 times the load caused by the acceleration imposed by the control rod drive mechanism. All spiders are tested in this manner.
3.
All rods are checked for integrity by the methods described in Section 4.2.4.2.3.
4.
To assure proper fitup with the fuel assembly, the rod cluster control, burnable poison and source assemblies are installed in the fuel assembly without restriction or binding in the dry condition. Also a straightness of 0.01 in./ft is required on the entire inserted length of each rod assembly.
The full length rod cluster control assemblies are functionally tested, following core loading but prior to criticality to 66 l
demonstrate reliable operation of the assemblies.
Each assembly 1
l is operated (and tripped) oe +4ma finw/ cnid ennd4+4nns a+
nn j
r l
l ene time 34 fn11 finw/cnla ennaitions, and era + % at full l
flow / hot conditions. Also any rod mechanism with a drop tine di l4 se & Y' l
outside the statistical two-sigma limit will,be. tr_ipped an
~
%rt t g
l additional s4w. times, at-the-plant conditier for dich the drop l
time was measured to be outside the two-sigme limit. Tim. sash rue not ran no.a t es+., w r
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P B R P R C R M C C F R C R P C t
T C U R P I
. to TXX-89139 March 22, 1989 CPSES/FSAR Page 27 of 35 Q423.2 Section 14.2.5 states that the JTG will review and approve the results of " required" tests before fuel loading.
If portions of any preoperational tests are intended to be conducted, or their results approved, after fuel loading:
(1) list each test; (2) state what portions of each test l
will be delayed until after fuel loading; (3) provide technical justification for delaying these portions; and (4) state when each test will be completed (key to operating modes defined in your technical specifications, or_to power ascension test power levels defined in Chapter 14).
Note that any test that you do not intend to begin prior to fuel loading should be included in your startup test phase instead of the preoperational test phase.
R423.2 (1)
Tests:
a.
Reactor Control System Test b.
Incore fiuclear Instrumentation Test (2)
Portions of Test Being Delayed a.
Reactor Control Systen Test The test summary table 14.2.2 (Sheet 41 of 60),
is to be modified to incorporate the following test method whose performan~ce will be delayed until af ter fuel loading.
"A functional demonstration at approximately 30% power to verify the Reactor Control System automatically maintains the proper reactor coolant average temperature conditions."
('
423-2 JULY 27, 1978
At'tachment 1 t'o TXX-89139 March 22, 1989-P, age 28 of 35 CPSES/FSAR I
b.
Incore Nuclear Instrumentation System Test i
Table 14.2.2, sheet 45 of 60, test method 2.
~
l Following core loading and insertion of the j
thimbles in the core, ensure free passage to I
all positions.
1 Table 14.2.2, sheet 45 of 60, test method 4.
l
_ During flux mapping, verify detector response l
to neutron flux and make final limit switch settings.
(3)
Justification i
a.
Reactor Control System Test L
The funtional test portion is being delayed 4
(
because the actual signals and' conditions required to perform the functional test cannot be obtained prior to fuel loading, b.
Incore Nuclear Instrumentation System Test Testing of the system for the items listed in (2)b above require the thimbles to be installed. The thimbles are installed into the core following completion of core loading
~
operation. The response of the detectors cannot be verified until the detector is placed i
in a neutron flux. Final limit switch settings will be completed when the thimbles are installed and the reactor coolant system is at
}
' 2ero power operating temperature.
(
~
423-3 JULY 27, 1978
%a
-. to TXX-891391 March 22,1989 -
CPSES/FSAR P, age 29 of 35 (4)
-Test Completion a.
Reactor Control 1 System Test The functional demonstration shall be completed while in mode 1 at a steady state power level so ot of approximately,36%.o.
b.
Incore Nuclear Instrumentation System The items listed in 2 ab'ove shall' be completed.
while the plant is in mod $ 2 at. a power level less than 5%.-
l
(
1 l
l l
l l
- l i
i I
423-4 JULY 27. 1978 i
l l
L___________._________
Attachment I to TXX-89139 CPSES/FSAR
- March 22,1989 9h4b.$k We could not conclude from our review of the startup test summaries in Table 14.2-3 that all of the tests will be
('
comprehensive. Therefore, clarify or expand the summaries to address the following:
_...J l
1.
Reactor Trip System Test - State your plans to demonstrate the proper operation of interlocks that prevent closing of both reactor trip breaker bypass breakers simultaneously.
2.
Effluent Monitoring Test - State your plans to also 1
demonstrate the proper performance of process and I
area radiation monitoring equipment under operating conditions.
Describe the portions of the test performed at initial fuel loading as shown in Figure 14.2-4.
I 4
3.
Control Rod Reactivity Worths Test - State how you will determine which RCCA is most reactive.
Clarify
(
the test method to show that the worth of all RCCA banks t 11 be measured.
4.
Loss of Offsite Power Test - State your plans to initiate the transient from an initial condition of generator output of at least 10 percent power.
The transient should be initiated by opening the generator output breakers in order to simulate a loss of offsite power. This test should demonstrate (for l
~ ^ ~
approximately 30 minutes) that the necessary
)
equipment, controls, and indication are available following the station blackout to remove decay heat from the core using only emergency power supplies.
5.
Rod Drop Tests - It appears that you do not intent to 423-35 MAY 31, 1979
ta h nt o TXX-89139 CPSES/FSAR P, age 31 of 35 conduct this test in accordance with Regulatory Guide
(
1.68 (November,1973) which includes drop time measurements of each rod at cold no-flow, hot no-flow, cold full-flow, and hot full-flow. Modify your test summary to show that the test will be conducted in accordance with the regulatory guide or provide technical justification for any exceptions.
Also describe the addditional drop tests that will be required for the fastest and slowest dropped rods and state whether these requirements apply to the fastest
~
and slowest rod at each test condition.
~
6.
Flux Distribution Measurements Test - Specify the control rod configurations for which flux maps will be obtained.
7.
Core performance Evaluation Test - Expand the test to include verification of calibration of flux and
{
temperature instrumentation (Regulatory Guide 1.68, Nov. 1973, Appendix A, Section 0.1 9).
8.
Remote Shutdown Test - Expand the test abstract to show that the test will be performed in accordance with Regulatory Guide 1.68.2, Revision 1, July 1978.
9.
Turbine Trip Test - The acceptance criteria for this test should be modified to 1) identify the parameters or variables to be monitored 2) provide assurance that the transient results will be compared with predicted results for the actual test case, and 3) provide quantitative acceptance criteria and their bases for the required degree of convergence of actual test results t<ith predicted results for the monitored variables and parameters.
(
423-36 MAY 31, 1979
Attachment I to TXX-89139
- March 22, 1989 CPSES/FSAR Page 32 of 35 R423.16 1.
The Reactor. Trip System Test Summary (Table 14.2-3,
(
sheet 6) has been expanded to test the reactor trip
)
bypass breaker interlocks.
\\
2a.
The performance of the process radiation monitoring equipment shall be demonstrated by comparison of monitor indication with the results of radiochemical analysis. Refer to the Process and Effluent Radiation Monitoring Test Summary.
2b.
iThe pec4=nca # the are r:diat'icr mea-i-tc. 2 ihell
~
_bc decuensti eted by cc pari:On ;f peniieneni crea monitor indication sith the indications of porth
^
a@**
oa a IMI su r vey insis us.6nt5 during pl6ni r adioi. ion surveys.
Refer to the Radieticr. Mor. iter and survey t est -
S mma ry.
2c.
The Process and Effluent Monitoring test shall begin
{
during the low power test phase to verify as early as possible and to_the extent practical the response of the process and effluent radiation monitors. Refer to the revised Initial Startup Test Schedule, figure 14.2-4.
3a.
The NSSS vendor will determine which RCCA is the most reactive.
3b.
The Control Rod eactivity Worths Test Summa.; (Table
/
14.2-3, sheet A )A dhas been revised to state selec that t.he worth of th control and shutdown banks 4
shall be measured.
(
423-37 MAY 31, 1979 l
1 G'r
.M e
r A.;.. '. to TXX-89139
~
March 22, 1989 Page 33 of 35.
iasexT Revise response to: question Q423.16.2b as follows:
R423.16.2b The. performance of area radiation monitors is satisfactorily f
demonstrated during the preoperational phase. The monitors are functionally checked and communications to the control terminal verified. The instrumentation department performs detector calibrations and provides a final check of the system operability.by exposing the monitor / detector to a source-of i
knovn activity and analyges the results.
During power operation the Radiation Protection department perform routine checks of the monitors by initiation of a check source and verifying system response.
These checks and measures provide sufficient testing of the area radiation monitors for operability assurance during power ascension.
o a
P%.
- W
$4, up l
I
{
Attachment I to-TXX-89139-CPSES/FSAR i
March 22, 1989 i
P, age 34 of 35 l
4.
The Loss of Offsite Power Test Summary (Table 14.2-3,
{
sheet 16 of 23), has been revised to state that the
~
generator output is at approximately 10%.
The transient shall be initiated oy a manual reactor
' i trip which will initiate a turbine-generator trip in order' to simulate a loss of turbine generator coincident with a loss of all offsite power.
IV 5.
The Rod Drop Test Summary (Table 14.2-3, sheet y[g PM has been revised to clarify.the plant conditions at the time of the tests and to describe the
. additional testing for the fastest and slowest dropped rods.
6.
Flux Distribution Measurement Test:
at/ rool.s 04 (M5 A Flux mapp shall be obtained at the fe!! cuing control
{
rodconfiguration/f l
1 a.
^ rods cut (fn,0h "u.
Cunt +^! henk 0 inserted-T.
Control b:rh, 0, C, S :t the hot zero pc.-cr i n3ei L i va Hett-,
d.
Dnandn rnd ejectica - Centrol bar' 0 et 5 h control benks C, 9 et the het zere power
_insertien limit end the predicted 'ighc:t "~th _
ejactad red grc:ter ther J0,0 s_tane-Unics5 5peciTied, all ether ;0ntr 1 ?nd chotdown _
banks emain w1Lndrawn fr0T the coactne enro (at 228.,
stonei-(
423-38 MAY 31, 1979
. to TXX-89139 CPSES/FSAR
- March 23,1989 Page 35 of 35 7.
The Core Performance Evaluation Test Summary, (Table 14.2-3, sheet 19 of 23), has been revised to include verification of calibration of flux and temperature instrumentation.
8.
Remote Shutdown Test Summary (Table 14.2-3, sheet 21 of 23), has been revised to conform with Regulatory Guide 1.68.2, Revision 1, July 1978.
m 9.
Turbine Trip Test Summary - Identification of variables to be monitored and quantitative acceptance criteria shall be specified in the detailed startup test procedures. Data obtained during the transient shall be analyzed and the results shall be compared with predicted results for the actual test case.
(,'
423-39 MAY 31, 1979
- to TXX-89139 March 22, 1989 Page 1 of 18 STARTUP PROGRAM REVISIONS OVERVIEW The startup test program modifications submitted in this document are identified below.
Details on each item are provided in the following sections.
For each case, the modification is described, a technical justification is provided, and precedence is identified, where applicable.
In addition, the necessary changes to ensure that the FSAR accurately reflects the PAT program are provided. The modifications include:
- 1. Deletion of the turbine trip from 50% RTP for Main Steam System-dynamic response testing.
- 2. Modification of the Reactor Coolant System Flow Coastdown Test acceptance criteria.
- 3. Deletion of the Area Radiation Monitoring check against the portable instrumentation during the radiation surveys.
- 4. Utilization of the " Rod Swap" methodology for Control Bank reactivity worth measurement.
- 5. Modification of the low power, Mode 2, power distribution measurement commitments to be consistent with the " Rod Swap" program.
- 6. Elimination of the Cold Full-Flow and Cold No-Flow rod drop timing tests.
- 7. Combination of the 30% and 50% RTP testing plateaus.
- 8. Elimination of the 90% RTP testing plateau.
- 9. Modification of the at power, Mode 1, power distribution measurement commitments to be consistent with the modified power l
plateaus.
- 10. Modification of the load swing testing commitments to be consistent 4
with the modified power plateaus.
- 11. Modification of the Remote Shutdown test to take credit for the cooldown portion being performed during the Hot Functional Testing per Regulatory Guide 1.68.2.
- 12. Modification of the Automatic Reactor Control System tests to be t
consistent with the modified power plateaus.
. Attachment 2 to TXX-89139 March 22, 1989 Page 2 of 18 ITEM:
PLANT TRIP FROM 50% RTP FOR MAIN STEAM DYNAMIC RESPONSE TESTING (FSARSection 3.98.2.1.2) l DESCRIPTION:
FSAR Section 3.98.2.1.2 commits to performing dynamic response testing of the Main Steam System during turbine trips from 50% and 100% RTP.
MODIFICATION:
The testing commitment at 50% will be deleted.
JUSTIFICATION:
RG-1.68 Appendix A, paragraphs 5.1.1 and 5.n.n relate to the dynamic response of the plant following a turbine trip at 100% RTP.
Paragraph 5.o.o relates to the verification by observations and measurements, as appropriate, that certain piping and comoonent movements, vibrations, and expansions are acceptable. RG-1.68 does not require any 50% RTP trip to meet the requirements for any of the above noted paragraphs.
FSAR Section 3.9B.2.1.2 states that the purpose of the dynamic transient response testing for the specified piping systems is to ensure that severe vibrations do not exist. The testing acceptance criteria for dynamic transients is based on the allowable design stress limits for occasional loads such that the induced stresses do not exceed the allowable stresses.
An evaluation performed by the Architect Engineer has concluded that the test at 50% provides little or no new information regarding the Main Steam System response to a turbine trip when considering the commitment to perform measurements during the 100% RTP trip.
Specifically, the 100% RTP trip bounds the 50% RTP trip.
Step and ramp load changes occur during the power ascension as part of the normal operation, therefore any adverse behavior of the Main Steam System would be observed.
PRECEDENCE:
No turbine trip tests were prescribed for Byron, Braidwood, McGuire, or Catawba plants (all 4-loop, similar vintage and power rating as CPSES) for' verification of proper steam piping response.
. Attachment 2 to TXX-89139 March 22, 1989 Page 3 of 18 l
ITEM:
REACTOR COOLANT SYSTEM FLOW C0ASTDOWN (FSAR Table 14.2-3, Sheet 3) j DESCRIPTION:
The acceptance criteria for the rate of reactor coolant flow decrease requires that it be slower than the rate of flow decrease assumed in the loss of flow analysis in the FSAR.
MODIFICATION:
The acceptance criteria is revised to state that the flow decrease must satisfy the Westinghouse provided acceptance criteria which validates the loss of flow analysis of the FSAR.
l JUSTIFICATION:
1 FSAR Figure 15.3-9 provides the normalized core coastdown flow vs. time l
for the four pump coastdown condition.
The figure provides flow rates I
which are conservative at each time interval with respect to actual flow rates.
The flow coastdown test data and FSAR Figure 15.3-9 data l
are not directly comparable because the initial conditions assumed for the Loss of Flow FSAR analysis are not the same as those used in the startup test (i.e., the FSAR assumes 100% RTP whereas the test is conducted at hot standby conditions).
In addition, the FSAR uses a I
conservative, normalized generic pump coastdown curve but the test l
measures actual core flow.
I 1
Westinghouse will provide acceptance criteria for the startup test conditions and actual core flows which correspond to the initial conditions and core flows assumed in the FSAR analysis. This will verify the conservatism of the FSAR analysis.
PRECEDENCE:
This methodology has been used by Westinghouse plants since 1985.
l l
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- Attachment 2 to TXX-89139 March 22, 1989 Page 4 of 18 ITEH:
l AREA RADIATION MONITORING AND RADIATION SURVEYS (FSAR Table 14.2-3, Sheet 12) 1 DESCRIPTION:
FSAR Table 14.2-3, Sheet 12 and the response to NRC Question 423.16 discuss, in part, verifying the performance of the area radiation l
monitoring equipment.
MODIFICATION-The referenced FSAR sections are revised by deleting the commitment to verify the performance of the area radiation monitoring equipment during the PAT program.
1 JUSTIFICATION:
RG-1.68 Appendix A, paragraph 1.k requires a preoperational test to I
demonstrate the proper operation of the area radiation monitoring i
equipment.
This item is captured in FSAR Table 14.2-2, Sheet 25.
j RG-1.68 has no similar item for the area radiation monitors during the j
power ascension phase.
j 1
In response to NRC Question 423.16, CPSES committed that
... performance of the area radiation monitors shall be demonstrated 1
by comparison of permanent area monitor indications with indications of portable survey instruments during radiation surveys". Whereas the i
portable area monitoring is being conducted during the power ascension phase to "... establish the adequacy of shielding and to identify high 1
radiation zones..." in accordance with RG-1.68 Appendix A, paragraph 5.b.b.
The comparison between this survey and the indications on the permanent monitors would not provide a meaningful demonstration of the performance of the area monitors. This is because good shielding design will, for the most part, cause the area radiation monitors to be reading in or below their minimum detectable ranges which is not particularly useful for the purpose of comparison. The effectiveness I
of the area radiation monitors is demonstrated during the preoperational test program and by routine periodic use of a calibrated check source. No additional verification during the power ascension testing phase is necessary.
PRECEDEN_Gli TV Electric is not aware of any plant performing the Area Radiation Monitoring Equipment check during the power ascension program.
- Attachment 2 to TXX-89139 March 22, 1989 Page 5 of 18 ITEM:
CONTROL R0D REACTIVITY WORTHS (FSAR Table 14.2-3, Sheet 15)
DESCRIPTION:
The current commitment as described in Table 14.2-3 is as follows:
- 1. The Boron Concentration Exchange method for the determination of the differential and integral rod worths of the Control and' Shutdown Rod Cluster Control Assembly (RCCA) banks is implicitly assumed.
- 2. The minimum boron concentration required to maintain the reactor shutdown with the most reactive RCCA stuck in the full out position is to be determined.
- 3. The acceptance criteria is that the total integral reactivity worth l
of all Control and Shutdown banks, less the highest stuck RCCA, is greater than or equal to the value used in the Safety Analysis.
MODIFICATION:
l The reference FSAR table is revised as follows:
- 1. Bank Exchange (Rod Swap) is the preferred method with Boron Concentration Exchange as the alternate method for determining the rod worths of the Control and Shutdown RCCA banks.
l
- 2. The minimum boron concentration will not be determined.
- 3. The acceptance criteria are revised to require that the rod worths 1
are determined to be within the design specifications.
(These is consistent with those noted in WCAP-9863-P-A.)
JUSTIFICATION:
l l
RG-1.68 Appendix A, paragraph 4.b states: "# measurements of control rod and control rod banks reactivity worths to (1) ensure that they are in accordance with design predictions and (2) confirm by analysis that the rod insertion ilmits will be adequate to ensure a shutdown margin consistent with accident analysis assumptions throughout core life, with the greatest worth rod stuck out of the core." Both the original and the revised versions of this test satisfy RG-1.68.
The purpose of this revision is to update the FSAR to the correct testing acceptance criteria and to allow additional flexibility in methodology selection.
The following provides an itemized justification:
l
' to TXX-89139 March 22, 1989 Page 6 of 18 l
CONTROL R00 REACTIVITY WORTHS (continued)
- 1. The Rod Swap. methodology uses the rod worth measurement technique described in WCAP-9863-P-A, " Rod Bank Worth Measurements Utilizing Bank Exchange", May 1982. This topical report has been reviewed and approved for initial fuel load and subsequent reload cycles in the Safety Evaluation Report transmitted to Westinghouse by NRC letter dated May 23, 1983, from Cecil 0. Thomas to E. P. Rahe, Jr.
- 2. The minimum boron concentration determination was a carryover from the old (vintage 1975) test methods.
The minimum boron concentration provides no meaningful information for plant operations or as a design verification parameter, therefore it is no longer measured per the Westinghouse guideline procedures.
- 3. Verification of the design by comparing the measurements to like predictions assures the accuracy of the design methods and assumptions and thus the adequacy of the required Shutdown Margin and Safety Analysis.
Review criteria are imposed on each RCCA bank measured and an acceptance criteria is imposed on the sum of all of the banks measured, per WCAP-9863-P-A.
PRECEDENCE:
Beaver Valley Unit 2 - Initial Cycle Startup.
Shearon Harris Unit 1 - Initial Cycle Startup.
_ to TXX-89139 March 22, 1989 Page 7 of 18 IIM:
l R0D DROP TESTS (FSAR Table 14.2-3, Sheet 19)
DESCRIPTION:
FSAR Section 4.2.4.3, FSAR Table 14.2-3, Sheet 19 and NRC Question 423.16 address rod drop testing.
The FSAR states that rod drop testing will be performed at cold no flow, cold full flow, and at hot full flow conditions. Also, the FSAR states that six additional rod drop tests will be performed for each rod whose measured drop time deviates from the mean by more than two standard deviations.
MODIFICATION:
-s-The cold rod drops, both no flow and full flow, are being deleted. Also, at least three additional rod drop tests will be performed for aach rod whose measured drop time deviates from the mean by more than two standard deviations.
JUSTIFICATION:
RG-1.68 Appendix A, paragraph 2.b states "...to the extent practical, testing should demonstrate control rod scram times both at hot zero power and cold temperature conditions, with flow and no-flow conditions in the reactor coolant system as required to bound conditions under which scram might be required".
For CPSES, by Technical Specifications, critical operations are only permitted with the plant hot and reactor coolant pumps operating. Also the Technical Specifications require only hot full flow testing.
The primary reason such testing is performed for PWRs is to cxercise the rods and to provide early detection of problems in the rod mechanical system prior to plant heatup.
Since the operability of the Rod Position Indication System and the Rod Drive System are tested prior to plant heatup, the intent of the cold drop tests is met by these other actions.
Data from Catawba Units 1 &2, and Farley Units 1 & 2 were reviewed and the cold rod drop time tests did not provide information unique and different than the results at the hot full flow condition.
Furthermore, this data shows that the Rod Drop Tests performed in the hot full flow condition are sufficient to detect any system problems. TV Electric considers the likelihood of a problem requiring a plant cool down for repair which is not detected prior to plant heatup to be acceptably low.
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- Attachment 2 to TXX-89139 March 22, 1989 Page 8 of 18 R0D DROP TESTS (continued)
The change from six additional drops to at least three additional drops on those rods whose measured drop times deviate from the mean by more than two standard deviations is to make the FSAR consistent with the test procedure and RG-1.68.
RG-1.68 Appendix A, paragraph 2.b recommends that at least three additional drop time measurements be made on those rods whose drop time deviates from the mean by more than two standard deviations. Three additional measurements are sufficient to verify the validity of the initial measurement.
This number is also sufficient to determine that the deviating rods are consistent or to demonstrate improvement.
PRECEDENCE:
Callaway, Byron Unit 2, Braidwood Units 1 & 2, and Vogtle Units 1 & 2 did not perform cold rod drop time testing and were not required to cool down for repairs to this system.
Byron Unit 1, Callaway, Catawba Units 1 & 2, and Hillstone Unit 3 performed three additional drops for those rods whose drop time deviated from the mean by more than two standard deviations.
- Attachment 2 to TXX-89139 March 22, 1989 Page 9 of 18 JTEM:
FLUX DISTRIBUTION MEASUREMENTS (FSAR Table 14.2-3, Sheet 20) l DESCRIPTION:
FSAR Table 14.2-3, Sheet 20 requires that the core flux distribution be measured at 3-5% RTP for various rod configurations.
The discussion of rod configuration is also provided in the response to NRC Question 423.16, Item 6.
MODIFICATION:
FSAR Table 14.2-3, Sheet 20 and the response to NRC Question 423.16 Item 6 are revised to:
- 1. Allow the use of more sensitive equipment to augment the installed plant equipment by revising the power level to "below 5% power".
- 2. Perform the measurement at only the "All Rods Out (AR0)" condition (i.e., delete the tests at the " Control Bank D inserted" and the
" Control Banks D, C, 8 at the hot zero power insertion limit" configurations).
JUSTIFICATION:
- 1. The previous power level of 3-5% RTP for the flux distribution measurement is correct if the installed plant equipment is not augmented with more sensitive equipment.
The utilization of more sensitive equipment will allow the measurement to occur as low as the "Zero Power Testing Range", but typically with this equipment the measurement is taken at the level of " Nuclear Heating".
Performing the measurement at this lower power level is consistent with RG-1.68 Appendix A, paragraph 4.e, i.e., it has no adverse impact on the data, and it provides some reduction in testing duration since the time to raise and lower power to the 3-5% level is no longer required.
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Attach' ment 2 to TXX-89139 March 22, 1989 Page 10 of 18 FLUX DISTRIBUTION MEASUREMENTS (continued) l
- 2) RG-1.68 Appendix A, paragraph 4.e states " Determination of flux distribution for comparison with distribution assumptions or
}
predictions to provide a check for potential errors in the loading
\\
or enrichment of fuel elements or lumped poison elements and to check for mispositioned or uncoupled control rods...".
This determination is best performed in the AR0 condition where the insertion of Control Rods does not interfere with the measurement / prediction relationship. Additional measurements at different rod positions were performed in conjunction with the 1
control rod worth measurements when these configurations were l
achieved during the testing process.
These data were primarily used
{
to aid the fuel designer in the resolution of a measurement criteria failure.
In previous plant startups when a significant criteria failure occurred, the AR0 flux distribution was repeated to augment l
the other measurements, where the other information was gsnerally I
not used.
Since the Rod Swap method is planned for CPSES, the i
configurations of " Control Bank D inserted" and " Control Banks l
D, C, B at the hot zero power insertion limit" will not be achieved.
I Sufficient information will be gathered from the performance of the l
bank worths via Rod Swap, with corresponding comparisons of measured vs. predicted bank positions, to verify the adequacy of the design models. Therefore, in light of the limited value of this additional information, CPSES will only perform additional Flux Distribution Measurements at the recommendation of the NSSS Vendor to aid in the resolution of measurement criteria failure, should such occur.
l PRECEDENCE:
I Beaver Valley Unit 2 and Shearon Harris Unit I as part of the startup program optimization utilizing Rod Swap.
Catawba Unit 2, Farley Unit 2, and McGuire Unit 2 using the first unit as a reference.
l i
1
- Attachment 2 to TXX-89139 March 22, 1989 Page 11 of 18 ITEM:
TESTING AT 30% RTP (FSAR Table 14.2-3, Sheets 22, 23, & 33) l' DESCRIPTION:
FSAR Section 14.2.5, FS.9 Figure 14.2-4, FSAR Table 14.2-3, Sheets 22, 23, and 33, and the response to NRC Question 423.2 discuss testing at 30% RTP.
In addition, FSAR Section IA(B) does not take exception to the implicit RG-1.68 requirement of performing certain tests at 25%
(30%) RTP.
MODIFICATION:
The above FSAR references are modified to replace the testing conducted at 30% RTP with testing at 50% RTP. Where testinq had been previously specified for both 30% and 50% RTP, only the 50% RTP test will be performed.
JUSTIFICATION:
The tests that are presently prescribed to be conducted at the 30% RTP plateau include the following:
Core Performance Evaluation.
Unit Load Transients (10% Step Load Change).
Automatic Reactor Control System Test.
The core performance data that could be obtained at 30% RTP is utilized for gross calibration adjustments of the Nuclear Instrumentation System (NIS) prior to power escalation to 50% RTP. This activity will be performed at 25-30% RTP as a hold prior to escalation to 50% RTP. The i
l flux distribution measurement at 30% RTP will not be performed unless the peaking factors measured at low power do not support escalation to 70% RTP, the NIS trip setpoint for the 50% RTP testing plateau. This is per the direction of RG-1.68 Appendix C, paragraph 4.h.
The unit load ' transient at 30% RTP was intended to be a precursor to the same transient test for higher power' levels, and the 50% load rejection transients.
However, a load change transient at 30% RTP does not provide useful predictions about whether acceptable control system performance would be obtained at the higher power levels. A step load q
decrease from 30% RTP results in the main feedwater control valves i
trying to operate at the lower acceptable end of their operating range, leading to poor control system response. Making control system adjustments (especially to steam generator level control) to address i
the operating concerns at 30% RTP has a high probability of causing more significant control problems at the higher power levels. This was noted during startup at such plants as Shearon Harris and Byron Unit 1, j
where setpoint adjustments made at the 25-35% RTP resulted in i
unacceptable performance during the 50% load rejection transients.
J The Automatic Reactor Control System test is intended as a precursor to the Unit Load Transient test and is performed at 30% RTP.
It is i
designed to ensure that the automatic rod control system can restore the Reactor Coolant System (RCS) temperature to within a 1.5 Deg-F deadband of the reference temperature.
Prior to 50% RTP, proper operation of this function would be demonstrated by observation during the normal power escalation, where the control rods will be in 1
automatic and already controlling the RCS temperature to within the deadband.
PRECEDENCE:
Vogtle Units 1 & 2, Farley Units 1 & 2 performed the nominal 25% RTP l
l testing at a nominal 35% RTP for stability considerations. No plant j
to date has combined the testing plateaus in this fashion.
{
ANSI /ANS-19.6.1-1985, " Reload Startup Physics Tests for Pressurized i
Water Reactors", requires a low power flux distribution at or below 30%
RTP for core loading verification.
The low power flux distribution measurement meets the intent of this standard.
- to TXX-89139 March 22, 1989 P, age 13 of 18 ITEM:
CORE PERFORMANCE EVALUATION (FSAR Table 14.2-3, Sheet 22)
DESCRIPTION:
l FSAR TABLE 14.2-3, Sheet 22, currently requires that Reactor Coolant System (RCS) parameters and incore data be recorded at five power lovels; 30%, 50%, 75%, 90% and 100% RTP.
The acceptance criteria states that core performance margins are within design predictions for normal and abnormal rod configurations.
MODIFICATION:
1.
The testing requirements at 30% RTP are modified.
Specifically, the flux distribution measurement is deleted.
2.
The testing requirements at 90% RTP are modified.
Specifically, the flux distribution measurement is deleted.
3.
The reference to abnormal rod configurations in the acceptance criteria is deleted.
JUSTIFICATION:
1.
The flux distribution measurement at 30% RTP is deleted based on the discussion under " Testing at 30% RTP".
2 The core performance data at 90% RTP is typically obtained for a calibration check of the NIS and Process Temperature Instrumentation Systems prior to power escalation to 100% RTP.
The flux distribution measurement at 90% RTP will not be performed unless the peaking factors measured at 75% RTP do not support escalation to 100% RTP per RG-1.68 Appendix C, paragraph 4.h.
Although testing at 90% is not required by RG-1.68, recording of RCS parameters will be performed at 90% RTP as a hold point prior to escalation to 100% RTP.
3.
The reference to abnormal rod configurations in the acceptance criteria is based on performing the pseudo rod ejection and drop tests at 30% and 50% RTP respectively.
These tests were deleted in a previcus FSAR amendment, but the reference in this section was inadvertently retained.
- to TXX-89139 March 22, 1989 Pa,ge 14 of 18 PRECEDENCE:
ANSI /ANS-19.6.1-1985, " Reload Startup Physics Tests for Pressurized Water Reactors", requires both an intermediate power distribution measurement between 40% and 75% RTP, and a power distribution at full power (greater than 90% RTP). The intermediate power distribution requirement is met by both the 50% and the 75% RTP core performance tests. The power distribution requirement at full power is met by the core performance test at 100% RTP. There is no explicit requirement for a 90% RTP test.
m_-___a,_---,
~ to TXX-89139 March 22, 1989 Page 15 of 18 ITEM:
UNIT LOAD TRANSIENTS (FSAR Table 14.2-3, Sheet 23)
DESCRIPTION:
FSAR Table 14.2-3, Sheet 23, requires that testing for transient response resulting from step changes in power level of 10% be conducted at 30%, 75%, and 100% RTP.
]
1 l
MODIFICATIONS:
The table is revised to change the 30% RTP test to 50% RTP and to i
delete the 75% RTP test.
JUSTIFICATION:
i RG-1.68 Appendix A, paragraph 5.h.h states " Demonstrate that the
{
dynamic response of the plant to the design load swings for the i
facility, including step and ramp load changes, is in accordance with i
design.
(25%,50%,75%,and100%)". The step load tests are preliminary tests performed prior to the more severe 50% Load Rejection i
l transients, and would verify proper controller response. The step load change transient would first be performed at the 50% RTP plateau.
i Performing this test again at 75% RTP does not result in additional l
information, as the control system responses (i.e., parameter variations from initial conditions and controller stability) are very similar for a step change transient initiated from either 50% or 75%
RTP. The 50% RTP plateau is high enough to allow for control system adjustments (largely in steam generator level control) that would still j
result in acceptable performance at higher power _ levels. The step load i
transient at 75% RTP also results in a smoother response than the same test at 50% RTP for the following reasons:
- The moderator temperature coefficient is more negative at the higher power level, resulting in smoother and more stable reactor power and temperature control.
)
i
- Feedwater temperature is higher at the higher power level, which acts as a dampening effect for steam generator level control.
The successful performance of this test at 50% and 100% RTP will bound the information at 75% RTP and verifies the proper operation over the operating range of interest.
In addition, the dynamic response of the plant is tested at 75% RTP by performing a 50% load reduction test.
This testing provides adequate confidence that the 10% load swing at 75% RTP need not be performed.
" to TXX-89139
' ~
March 22, 1989
.Page 16 of 18 i
UNIT LOAD TRANSIENTS (continued)
PRECEDENCE:
The plateaus at which plants have performed the transient tests have varied.
The Byron /Braidwood and McGuire plants have included in their FSAR 30-35%, 75%, and 100% RTP plateaus for performing the load swings, with plateaus of 75% and 100% RTP on Byron /Braidwood for performing a 50% load rejection and a plateau of 100% RTP on McGuire for performing a Loss of Electrical Load (essentially a 100% load rejection).
l Shearon Harris has listed plateaus of 30'/, and 75% RTP for performing a load swing and a single plateau of 75% RTP for I
performing a large load rejection; there were no commitments for doing transieat testing at 100% RTP.
l 1
l w___-_____-_-________-___-___________________-__________
- Attachment 2 to TXX-89139 March 22, 1989
,Page 17 of 18 ITEM:
REMOTE SHUTDOWN (FSAR Table 14.2-3, Sheet 25)
DESCRIPTION:
FSAR Table 14.2-3, Sheet 25, commits to demonstrate during initial startup testing, the capability of performing a safe plant shutdown, maintain the plant in a hot standby condition, and demonstrate the ability to cooldown from hot standby to cold shutdown conditions (300 Deg-F) from outside the control room.
MODIFICATION:
Table 14.2-3, Sheet 25, is revised to allow the portion relating to plant cooldown to cold shutdown conditions to be performed during the hot functional testing (HFT), preoperational testing, or initial startup testing (ISU), provided that the initial conditions required q
for test performance are satisfied.
JUSTIFICATION:
Regulatory Guide 1.68.2, Revision 1, paragraph C.4 states: "The demonstration of cold shutdown capability need not necessarily be i
performed immediately following the demonstration of achieving and maintaining safe hot standby from outside the control room. Rather, this cooldown portion of the test may be combined with another startup test requiring the reactor to be cooled down, as long as the procedures and acceptance criteria for the combined test meet all the elements of each individual test."
The cooldown, from 557 to 300 Deg-F, can properly be demonstrated r
during the HFT program, or the ISU program when the prerequisites are satisfied.
There are several tests currently planned during the HFT l
and ISU programs which meet the prerequisites and have the need to I
establish cold shutdown conditions following the test. The remote shutdown cooldown demonstration will be performed following one of these tests.
I I
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- Attachment 2 to TXX-89139 March 22, 1989
,P;ge 18 of 18 ITEM:
AUTOMATIC REACTOR CONTROL SYSTEM (FSAR Table 14.2-3, Sheet 33) l l
DESCRIPTION:
FSAR Table 14.2-3, Sheet 33, requires that the Automatic Reactor Control System (ARCS) demonstrate the ability to achieve and maintain proper reactor coolant temperature at 30% and at 75% RTP.
MODIFICATION:
FSAR Table 14.2-3, Sheet 33, is revised to perform the 30% test at 50%
RTP and to delets the 75% RTP test.
JUSTIFICATION:
The justification for performing the test at 50% vs. 30% RTP is provided in the discussion under " Testing at 30% RTP".
The Westinghouse NSSS startup guideline procedures only schedule this test at the 30% RTP plateau. The performance of this test at 50% RTP is consistent with the intention of these procedures and has been recognized by Westinghouse as an improvement for control system adjustment at CPSES. The test at 75% RTP is considered redundant to the 50% RTP test, therefore this test is deleted.
PRECEDENCE:
This test is usually performed only once during a plant startup, and the behavior of the ARCS is monitored during the unit load transient tests.
This is documented in the startup reports and FSAR's for McGuire, Catawba, Byron /Braidwood, and other plants.