NRC-88-0212, Suppl 9 to Interim Startup Test Rept
| ML20154L596 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 09/10/1988 |
| From: | Sylvia B DETROIT EDISON CO. |
| To: | |
| References | |
| CON-NRC-88-0212, CON-NRC-88-212 NUDOCS 8809260227 | |
| Download: ML20154L596 (201) | |
Text
,
THE DIrfROIT EDISON COMPANY r
j FIERMI 2 j NUCLEAR POWER PLAVf a
IlrTERIM STARTUP TEST REPORT i
SUPPLDIENT NO. 9 September 10, 1988 4
k 4 ,
l 4 ;
1 t L
l i'
1 1
hhk'E00227880920 P ADOCK 05000343 PNU 1
,1 a -r .- , ,-n -- -e,,, - - , - - -r - - - - - -- - - - , n-- ---- - ,
3 l
INTERIM STARTUP TEST REPORT Supplement #9 Dated 09/10/88 l
Revision Summary Forward Page General Update Page 1-3 Reference Addition 1 Pages 2-3, 2-4, 2-5, 2.6 Status Update Page 3 1-9 Status Update Pages 3 2-1, Test Results Update 3 2-4 through 3 2-42 (excluding illustrations)
Page 3 8-2 Test Results Update Page 3 9-3 Test Results Update Page 3 10-3 Test Rer:1ts Update Pages 3 11-9, 3 11-10, 3 11-11 Test Results Update Page 3 12-4 Status Update l
Pages 3 13-6, 3 13-7 Test Results Update l
Page 3 14-2 Test Results Update 1
Pages 3 15-3, 3 15-4 Test Results Update Pages 3 17-3, 3 17-4 Test Results Update Page 3 20-5 Status Update Pages 3 21-5, 3 21-6, 3 21-7 Test Results Updats Pages 3 22-1, 3 22-2 Test Results Update Page 3 23-3 Test Rer21ts Update Page 3 29-3 Status Update
IlffERIM STARUlf TEST REPORT supplement #9 Dated 09/10/88 Revision suasary Page 3 26-3 status Update Pages 3 27-6 through 3 27-8 Test Results Update Pages 3 28-2, 3 28-3 Test Results Update Pages 3 30-4, 3 30-5 Test Results Update
<- Page 3 31-7 through 3 31-10 Test Results Update Page 3 32-2 Test Results Update Page 3 33-1 status Update Pages 3 34-2, 3 34-3, 3 34-4 Test Results Update a
8 1
L l
l T
I
Supplement 9 FOREWARD This Supplementary Startup Test Report includes testing performed since Supplement #7 dated March 10, 1988 was transmitted to the NRC via NRC-88-0038 dated March 20, 1988. Letter NRC-88-0136 dated June 20, 1988 was transmitted to the NRC to serve as Supplement #8.
There was no update at that time since the plant had been in a Local Leak Rate Testing Outage from February 1988 through May 1988, and as of that letter, plant power operation had not yet reached Test Condition Six power / flow requirements.
Since the plant's return to power operation, thirteen of the required twenty six Test Condition Six tests have been completed as well as all four of the required Test Condition Four tests. In addition, the balance of the feedwater system tuneup and dynamic response testing not possible during Test Condition Three has been successfully completed. This discrepant testing was identified as Inspection Report Open Item 341-88003-01 contained in Inspection Report No.
50-341/87046 (DRP).
In this supplement we are transmitting an updated copy of the entire test report. Revision bars have been added to show where changes have been made, except for changes which are only cosmetic in nature or which only involve renumbering sections or pages.
The results sections of this report will be filled in as the tests ar e completed in the future.
Supples;nt 9 FERMI 2 FUCLEAR POWER PLAlfr INTERIM STARTUP TEST REPORT INDEI 1.0 Introduction 1.1 Purpose 1.2 Test Report Format 13 Plant Description 1.4 Startup Test Program Description 1.5 References 2.0 General Test Program Information 2.1 Chronology of Major Events 2.2 Matrix of Test Completion Dates 30 Test Results Summary 31 Chemical and Radiochemical 32 Radiation Measurements Fuel Loading 3.3 34 Full Core Shutdown Margin 30 Control Rod Drive System 36 SRM Performance and Control Rod Sequence 37 Water Livel Heasurements 3.8 IRH Performance 39 LPRM Calibration 3 10 APRM Calibration 1 3 11 Process Computer 3 12 RCIC 3 13 KPCI 3 14 Selseted Process Temperatures 3 15 Systen Cxperision ,
3.16 (Deleted) i 3 17 Cors forforsance l 3,18 (Deleted) 1 j 3 1-) (Deleted) .
4 3 20 Presrure Pegulator l
'.21 Feedwater System J ,
3 22 Turoine Ve,1ve Surveillance 3 23 MSIV 3 24 Relief Valves i 3 25 Turbine Stop Valve and Control Valve Fast Closure i
Supplement 9 FERMI 2 NUCLEAR POWER PLANT IlfrERIN STARTUP TEST REPORT INDEI 30 Test Results Suasary (Continued) 3 26 Shutdown from Outside Control Room 3 27 Flow Control 3 28 Recirculation System 3 29 Loss of Turbine-Generator and Ortsite Power 3 30 Stuady-State vibration 3 31 Reciro. Systen Flo.: Calibration 3 32 Reactor Water Cleanup System 3 33 Residual Heat Removal System 3 34 Piping Systen Dynamic Response Testing i
i d
_ . __.___ -,,_.__ , -___,~_-,._y -
_ _ _ , _ _ _ _ , - _ _---m.-y- y , _-_ - . _ .- ...,,.--_.-,,--__-------__yy _
SupplCZ nt 9 P:go 1-1 FERMI 2 NUCLEAR POWER PLANT IFTERIM STAR'11JP TEST REPORT 1.0 Introduction 1.1 Purpose h e purpose of this Interim Startup Test Report and its associated supplements is to provide a suasary of the test results obtained in startup testing completed from initial fuel load to the present. Bis report of plant startup and power ascension testing is submitted as required per Technical Specification 6 9 1.1. This interia report plus its supplements cover all testing applicable to the test conditions completed as described in UFSAR Subsection 14.1.4.8. Supplements will be issued as the remaining testing is completed, at the intervals specified per Technical Specification 6.913 Included in this report are descriptions of the measured values of the operating conditions and characteristics obtained during the test program and any corrective actions that were required to obtain satisfactory operation.
1.2 Test Report Format Sections 1.0 and 2.0 of this report provide general information about the Fermi 2 plant and the testing program. Section 3 0 provides a basic description of the testing we have performed along with a summary of the results and analysis obtained from each test. Each test summary is divided into three subsections covering the purpose, test criteria, and results of each test.
13 Plant Description The Fermi 2 Nuclear Power Plant is located in Frenchtown Township, Honroe County, Michigan. We Nuclear Steam Supply System consists of a General Electric BWR 4 nuclear reactor rated at 3292 MWL, coupled to an English Electric Turbine / Generator rated at 1100 MWe, constructed in a Mark I containment with a toroidal suppression pool.
4 his plant is owned and operated by the Detroit Edison Company and the Wolverine Power Cooperative, Incorporated.
i
Supples:nt 9 Pag) 1-2 1.4 Startup Test Program Description The Startup Test Phase began with preparation for fuel loading and will extend to the completion of the warranty demonstration. This phase is subdivided into four parts:
- 1. Fuel Loading and Open Vessel Tests
- 2. Initial heatup 3 Power tests
- 4. Warranty demonstration
'Ihe Startup Test Phase and all associated testing activities adhere closely to NRC Regulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors."
The overall objectives of the Startup Test Phase are as follows:
- 1. To achieve an orderly and safe initial core loading
- 2. To perform all testing and measurements necessary to dstermine that the approach to initial criticality and the subsequent power ascension are accomplished safely and orderly 3 To conduct low-power physics tests sufficient to ensure that physics design parameters have been met
- 4. To conduct initial heatup and hot functional testing so that hot integrated operation of specified systems are shown to meet design specifications 5 To conduct an orderly and safe Power Ascension Program, with requisite physics and systen testing, to ensure that when operating at power, the plant meets design intent
- 6. To conduct a successful warranty demonstration progran Tests conducted during the Startup Test Phase consist of Major Plant Transients and Stability Tests. The remainder of tests are directed toward demonstrating correct performance of the nuclear boiler and numerous auxiliary plant systems while at power. Certain tests may be identified with more than one part of the Startup Test Phase. Figure 1-1 shows a general view of the Startup Test Phase Program and should be considered in conjunction with l
l l
Supple: Int 9 P ge 1-3 Figure 1-2 which shows, graphically, the various test areas as a function of core thermal power and flow. Note that Figure 1-1 has been modified to reflect certain tests which we presently intend to delete from the Startup Test Pregram, as discussed further in Reference 1.5 3 For a more comprehensive description of the testing program refer to Reference 1.5.2.
1.5 References The following is a list of documents that provide supplementary information of the Fermi 2 Startup Test Phase Program:
- 1. Fermi 2 Technical Specifications, Se tion 6.
- 2. Updated Final Safety Analysis Repor;, Fermi 2 Nuclear Power Plant, Section 14.
3 Letter VP-86-0141, "Startup Test Program Changes", dated October 17, 1986, from Frank F.. A gosti to James G.
Keppler.
- 4. Letter NRC-87-0179, "Initial Test Program Changes" dated September 30, 1987, from B. R. Sylvia to U. S. Nuclear Regulatory Commission, Washington, D.C.
5 Letter NRC-88-0181, "Change in Startup Test Program",
dated July 14, 1988, from B. R. Sylvia to U. S. Nuclear Regulatory Connossion, Washington, D.C.
F-r I
FIGU8tE t-t STAst1UP TEST resOGstam Ope. T..t cas.3 6 t 6 .
v.w. B _ . - -
v en c. 4
- " ***. Ta'- ""'T .'_ _2_ 3_ . _.** _S . *__ "u.a.em.1 e a sc.R a e.atas ec.I s a a a a a a m.a a .t a m.~ . u a a : a
- a r
- 8 as , a e r.tR c . sm.=*-. n..,s. x l
s cae a a a :
s san e.e e.e c. a.a c e..a e.e : ; _ m a r e., a a a. . i B ee e.e sw c. a u e, t. ann c.n em.e a = x a m a se ar.a casame.it e a a m x a m
. . _ _ . _ _ . _ . . _ .. __. ~ _ __.. . . ... _ _ _ _ _ _
aa ev.c e- r a w a ac a a a cs., coe. s am 8 c a.,sv.s x n n as ma,m e . c R e s.).ce 4 s r.. a a se s. .ce.a er.c . v cu.e a z x as s,.e sa m a s e a
__ . ___ _.. _ . ~ . . _ . - ~ . __. . _ ___ _ . .
is se.t.e.as av ca. rusw c. s a a a a a Be to.l.a.as 39 (B. tee.49 se e we. m.,.am _ s.e cw . a.ne a n a n m.coy.4.=m.,.t...e =.ne = = =
2 r a e= s,.t - c a . e ,rese ntset
. .e 8 , a s.e ,.a.: cww.. x a a a se n
-u. .,t . wt
-sen.a a ecepeansaaay ud as v =4 v.s sw ata c ,s. se a..f.se as .est v. - e.e v.s as mm .se s1 xe
- retB B 4.e& M.Suk ae at.e v.s m =6 as ve n seo, v.s ac ..a v.a r..e ca e. re sp. a.sel se.so n .
- se s ee-. e. om.t4. com..a m esoia* x.
27 r8.ms c e 4 se 4 frt se, e.cere.neea sr.e *=. r asse
- tee- s,.r. e e.r e c. m a a m
- ==.co.se.et v.rarse.ea =
29 8 .f T-4 off.te. s%es.e . 1s.sDa 4 56me.e4 se.e e. 4 W X X 33 a.c6. ,. tee 4 sy.e rte. Ce t the.t 4.se R I si s ce., ci- sr a a 99 a. 44 I um..e me a Srse 3 I De Pept.g sy.e Dy 6c m y e a E E E 5.yo se . t f t c e .t o.a. aet g s.y 3 e ,( e go c.s.t e.l s SP * .c e. pe..lb4 e at ya sD = .Cre. d.flatt.s 9P
- STP... t.* v.*P *-
- s et,.e. se a-2 to, e e e die t e.,6 p. hy.gg gge g g t..
- r e E., se e.e.4 etee.a.e B see pesce.e a e 8 4 c b. +>=. t e.e T..e c d a t t 2 s 3.
- e. tes tee 6.e t.. ca.aae t 3. 3.s eens. syy... wa c p.me t te r fe.ete e T. c ate s a e a 3 t a s a. s ea et e.= telt se.tv s. t.t t .r teetta./g eetoe trap .ce e. et b.t de v..e on.aats s e.a 6. 7e puc e.a noe p.ec e et me. th.. ap .a t.e ehte an.e .a .s , .sen e .cre.. . a se ,p it ,e tF.4. .ho that e. e .Jttth.1t. ca...b. .r t. tee ite.polet.a t the tream. et p e t y.t.e. magm pn . c 44et .
Seep.ee e.t et s.-.a e.a.. Br.e s . v... S. e t.a a f e .t a conia.a rea. 6. ca. 3 ct 8 .t en th . .c .
emme .e.se4 cu.aaog . s. - -e..et .
Suppieenent 9 Page I-4 l
l .. , - - . - - - - , . ._ . ,
Supplement 9 Pego 1-5 FIGURE 1-2 APPRDIIMATE POWER FLOW MAP SHOWING STARTUP TEST 00WDITIONS 110 - - 3 i I ' ' i i i i i i -
A. hwret serewleteen g j gg _ ,,
B. Menemum retirasteten pwmp speed C. Anotytisoi se,we Hmet et mente, pe==r % eentrei TC6 90 - - D. Anotyt.ee r v , s.mn et me,te, ,,,we ,
h seatrei j go'_ _ Region il + _
TC 5
- e. 70 - -
f N
- g 1
{ 60 - ~ g #1 i e i 50 - -
TC 4 40 - -
p -
Y TC2 30 -- _
Minimum Power Line 20 -- -
h TC 1 Cavitation 10 - - e Region Typical Startup Path -
Region ill 0 f ' ' l ' i f i i i i 4 4 -i , i i i 0 10 20 30 40 50 60 70 80 90 100 110 Percentage of Core Flaw 3 notest
- 1. See Figure 1-1 for startup test titles.
- 2. Power in percentage of rated thermal Power 3292 NWT.
3 Core flow gn percentage of rated core recirculation flow.
100.0 x 10 lb/hr.
- 4. TC : test condition.
I
Supplement 9 Pag 3 2-1 2.0 General Test Program Infora.ition 2.1 Otronology of Major Events Date Received (5%) Facility Operating 03/20/85 License No. NPF-33 Started Fuel Loading 03/20/85 Completed Fuel Loading 04/04/85 Completed Open Vessel Testing 06/01/85 Initial Criticality 06/21/85 Received (Full Power) Facility 07/15/85 Operating License NPF-43 Com,.*1ted Initial Turbine Roll 09/26/85 Bypass Line Replacement / 10/10/85 Environmental Qualification Equipment Upgrade Outage Begins Neutron Source Changeout Complete 05/12/86 Outage Ends 07/24/86 Reactor Restarted 08/04/86 Completed Test Condition Heatup 09/03/86 Entered Test Condition One 09/1C/86 Initial Synchronization to Grid 09/?1/86 Condenser Repair Outage Begins 11/08/86 Reactor Restarted 12/18/86 Completed Test Condition % e 01/07/87 f
! Main Steam Line Instrument Tap 01/09/87 Repair Outage Begins Reactor Restarted 01/24/87 Entered Test Condition Two 02/24/87 i
Completed Test Condition Two 03/16/87 with Loss of Offsite Power Test
Supplement 9 PCge 2-2 mronology of Major Events (Continued)
Date hSR Refit Outage Begins 03/16/87 Reactor Restarted 04/03/87 Main Stsas Line Tap Repair 04/12/87 Outage Begina Reactor Restarted 05/10/87 South RFPT Damaged 05/13/87 Reactor Restarted 05/14/87 Commenced Test Condition Three 06/10/87 Testing Completed Core Flow Calibrstion 06/14/67 at 50% Power Outage to Repair Reactor Reciro 06/25/87 NG Set "B" Reactor Restarted 06/28/87 South Reactor Feedpump Returned 07/02/87 to Service Outage to Repair Feedwater 07/31/87 Check Valve Begins Reactor Hestarted 10/09/87 Commenced Test Condition Three 10/14/87 HPCI Test Sequence Completed Test Condition Three 10/24/87 KPCI Test Sequence NRC Authorization to Exceed 50% 12/05/87 Power Received Resumed Test Condition Three 12/09/87 Testing at 50-75% Power Completed Core Flow Calibrwtion 12/17/87 at 71% Power Completed Test Condition Three 12/26/87 Testing
/
Supplement 9 Pag) 2-3 Chronology of Major svents (Continued)
Date Commenced Test Condition Five 12/29/87 Testing Completed Test Conditi.1 Five 12/30/87 :
Testing Outage to Investigate Reactor 12/31/87 Feedpump Control problems Reactor Restarted 01/08/88 NRC Authorization to Exceed 01/15/88 75% Power Received Started 100 Hour Commercial 01/19/88 Operation Demonstration Run at
> 90% Net Generation Completed Commercial Operation 01/23/88 Demonstration Run Completed Core Flow Calibration 01/28/88 at 955 Power Scheduled LLRT Outage Begins 02/27/88 Reactor Rertarted 05/05/88 ;
Reactor Manually Scrammed Due to 05/07/88 .
Loss of 120 KV Bus 101 l Reactor Restarted 05/08/88 Reactor Scram During HPCI 05/08/88 '
Surveillance !
Reactor Restarted 05/09/88 !
Reactor Scram Due to High RPV 05/10/88 :
Pressure After Failure of l Turbine Bypass Valves Reactor Restarted 05/12/88 Commenced Test Condition Six 07/11/88 Testing (
I Completed Core Flow Calibration 07/17/88 '
at 975 Power i
t
Supplement 9 Page 2-4 Chronology of Major Events (Continued)
Date Outage to Repair Unidentified 0?/23/P3 Drywell Leakage 88-02 Reactor Restarted 08/06/88 Reactor Scras Due to False 08/13/88 Turbine Viaration Signal Reactor Restarted 08/14/88 e
Completed Feedwater Tuneup / 08/16/88 Dynamic Response Testing Performad One Recirculation Pump 08/20/88 Trip Test From Test Condition Six Commenced and Completed 08/20/88 Test Condition Four '.'esting i Plant Shutdown to Repaie 08/21/88 "B" Reactor Recirculation rump Discharge Valve Reactor Restarted 08/23/88 Plant Shutdown to Repair 38/29/88 "B" Reactor Recirculation Pump Discharge Valve l
i f
l l
Il
l l
- l ea l C
e eee eee e e e @ l e eee ee e e ee t
% %%% %%% % %% N
- e W M A A W A A A *= N N A @ O *= A @A WW@WA ==
, W= W *= A A W A A A es e=
- A O N a= A .= A W = w W A g$ e i
>M W % A A W A A A % % % A % % % A %A W % % W \ E l h h h h eeh h ee 3 e o O O O O O O O O O O a l
Pm Pm h P= h h h e ee e
eee
$ ** A A A A A A A A @ N s=A A A A A @ A @ @ O A A WD N A A A A A A A A N A A A ~ A A N A h N f4 A A M* % A A A A A A 4 A % % A A A A A % A % % % A A b N N
- N N N
- a= e- N= e= s=
4 e e e
- e e e e
% l d A A A A A A A A A A A A A O A A O A A A A A A [
U A A A A A A A A A A A A A N A A N A #, A A A A ,
P A A A A A A A A A A A A A % A A %A 4 A A A A b e e y
C O
% h P= Pm P= P. N N h P= i tot e eee e e e e e r 4 g %% %%% % % %% % .
w $ P* A A A A . A A N M @ A T A A A M A e @ A P= A (
Wb o *= A A A A A A N N N A N A A A N A == N A *= A
>K % % A A A A A A % % % A % A A A % A % % A % A M h N @ $ N O @ N N N ,
O a O O *
- O * * * ,
' h k h @ h P. >= P.
e ee e e U e e N %%% % %% % *
$ A 9 A A A A A A A O 89 A A A A O A O 9 A A *= "
W5 A O A A A A A A A * , O A A A A = A e O A A =
4
>> A % A A A A % A A % % % A A A A %A 09
% % A A %
M >=
M M N ** f9 O O O
- O O O O
@@ @@ h P= P= P= @@ @ P= @@ @
- ee e e eeeee a e e as aa e D l %% %% % % % % % % % % %% % 9 ,
! P= P= A A M @ A Pm f9 T N @ @ A T A T A a= am A e A ** t d W - N == A A N O A O O O O == N A O A O A N N A O A 9 t
% A % % A % A =
> % % A A % % A % % % % % % A % a O 0 OO * = * = O f4 - = OO O
- * * = 0 O O O * *=
- C * = a i
u !
ee o to e e@ ee @@@@ c ec w
& W e e e e e e ee e ee e e e a 3 %% %%% %%%% %%%% % %% 9 l
- @ T A O @ @ O 9 4
- A * = *= @ A A A A @ A N f9 > I
= A
- N f4 O N O A O O O O A A A e O A == Ce *
% % A % % % % % % % A w % % % A A A A % A % %
P=
- T @ e e@ @ e @ @ me >= h P= 0 '
5 O '! O O O O O O O O O O O O O O C b . =
, = . . . e 4 e e e Ig w%
A @ T A @ A A A A A A A A A N A A A A A A A A *
- I I
A A A A A A A A a= A A A A A A A A C 6 y y A ** O A O A A A A A A O
- 9
> A % % A % A A A A A A A A A % A A A
= C 3 e 1
g 9 9 T
- O k OO O O . .O g e i
s'l J V w a f
= @d @ C
- 30 0 3 9 ee e O V
- a !
3 %% % W C P= v t
b T T A A A A A A A A O A A A A A A A A A A A 4 0 e ut g
- N A A A A A A A A M A A A A A A A A A A A A wW@@ ,
i 9 eJ A A A A A A A A % A A A A A A A A A A A A W WT 2 y %. %. o e w b a oo O
- 8 4 9 e i 0
- M M I
- I e 8 e e C auu !
4 e w * > > +
3 e a 3 - L C g
= 0 3 e C
- C C ,
- v - *
- f i C
- e C b S w - * # e e C 6 -
- - O C
- L C e e D 0 V *
- G e 6 0 i O E = g
- 4 9 4
- e v .
i e em s E
- b h h b
= I C 3 0 0 6 0 0 u V l
> 4 3
6 >
- e *
- 0 C C 4
$
- 8 e 3
- O L b d 6 k W == D b o e C C 4 $ C 4 9 ** $ 3 9 8 $ $
es e 3 D DD e 4 0 0 9 d O w V D e e M aD G E w = a o w wy e y = w e C * = w u
> e 3 CC E C e. w w es .W O 3 e e e g 6 m m 8 D g S Z e6 3 e e O C Qa ) == $ > > e a C =
> w C a C= e E3 = a k k b e D D 4 W = 3 8 e b & 8 b D a & a4 6 0 6 a g e - O O e =e
- e C D G OO 3 0 *
- O = t*
- 9 <.* b > e w C 3 3
'1 O > ~ O e b *A e * = = 0 hA 3 0 6 3 9 86 9* 9 > * *
- e
- O O 6 a b e e d 9 O- 8 O- O
- v V a u w .4 W DD e vV e
- E &9 4 &9 3 4 C
- C d e e
- - e aO 4 4 0 W e O O e s = e - 6 b - - - a e 3 3 v u == 6
- tw e6 6tw a eD a4 w>== -= * * **
- 1 w D e = 0 3
- 3 a qs O == v = a w * *
- j 9 3 0 8 8 0 # D 4 3 3 4 2 # $ 4 a b U $ h O OO 4 eW W WW & w > 3 a > 2 > >
W 5 m. tb V e b == .J e & 3 I
5 e a e
~
h . e-N M T . .
i N.
N >
el- P. M , e@ ~ e e-------
O . h M W o@ - - N N N N N .#4 WA A. .
f i
l 1
1 l
4
---m -, ,- r= ---,,,-vw,.,m,- ,, c . , . - - -.---v.,--n,,n~r,---- ,,,-e, -~--n-- s v -- -
w
l 1 llI il!l 1 lI:I, l 9 _
t
- n 8 e6 8 m-
/ e2 s
- S S S > S 8 > 5 S l p e Ct
- S S S > S 2 > 5 S p g T5
- S S S > S / >8 1 S t n
e u a 0
t S P r
e ,
v d _
e > > > > > > > > > > e Cw > > > > > > > > > > n Tl > > > > > > > > > > S F
n w
e 8 b r 8 8 o
/ / f _
r > > > 0 > > > 0 > > t Cu > > > 2 > > > 2 > > i To > > > / > > > / > >
F 8 0
8 0
d e
r C
7 7 7 7 7 8 8 8 8 8 n ,
s / / / / / e e > >
6 5 > 8 4 > > 4
> 1 > >
k a
Cr > 2 1 1 1 Th > > / / > / / > > / t T 2 6 2 6 0 1 0 1 0 1 e v _
a h
7 8
7 7 7 7 8 8 8 8 "
?
/ / 2/ /
e 6
/
> 8 8 7 > > = C e.
. Ce 1 > 0 1 0 > > > l TT / / / > > > /
e 3 / 3.
3 3 3 3 t r
~
0 0 0 0 0 0 o
p L
6 8
6 8 d
. r e _
/ / e 8 e > 3 > > > 4 > > > > t 8 s Cn > 2 > > > 0 > > > > e / i TO > / > > > / > > > > l 7 h 2
+
0 0 p 1 t 1 1 m /
o 7 f .
i_ C 0 o _
_ 5 5 5 5 d 5 p 8 8 9 8 t e n u / / / /
y a1 t > > > > > 3 > 4 1 4 x
e e >
> > > 0 >
> > > / > / / /
1 3 2 t 7 n t
e 7 7 8 9 e 8 o _
2 0 0 0 0 e / l 7 t t 9 t 4
ne ee l
ps > > > > > > > > > >
b u
/ e .
2 S 7 1
n/
8 __
> > > > > > > > o . onI n 1 e OW >> > > > > > > > > > i n 5 /
3
) t od 2 _
m d i i ed 8 e d t mn _
u l ni r a n L
n e od o o i ud C nf L _
t Fa > > > > > > > > > > o r P a - e > > > > > > > > > > t C e sT s p r C o eL > > > > > > > > > > e t C r s s be%
( P t e n m8 .
s n s t o u .
e t o e i i N 4 s h s t 7 t e l i a r o t mm n a te e op t i a e h r C m D t r i t b n o n t r t 4
n o n b s s s t i e r _
o r l f f e C i y y e d o a t
. t C l S S R ef C f p -
t d s a e.
l e
p ne e arul o m e
r t C pl C n
- o. e m n v a ual m
r y e R r o r e e i t
es s o m o a o a f f l h C
m o
l e vat l l u t
sy P el e m n e y r F l r s f e
p e e a s
t l
r t
o _
t aC O S e u C R D c r a _
t i V t sm e e o e r -
s f t m nt a e r t m b n C t e e ps o l o s e t te ae te n n T
f t
s e
oa r tF f S t oif m s r t f aO nS w y a H s y t o gl n a d g a e __
o T e n nt o s S i n r /
o o e g t
~
ev wm nl ooc o uf i . r a C ot C o u g n si C n J s i e t C i i ia do r a r t d n t t i a b _
e bV t R w t s r t C i i s d r t r u o c s b C s s p e o d n u N
a T
u S
h l e oi e e e F R L V R p R P i
T N A I t _
= a * * .
t se 5 6 7 8 9 0 8 2 3 4 S > *
- 2, eM 2 2 2 2 2 3 3 3 3 3 S >
- S >
- 2 T
~
L 4
Supplement 9 PCge 301-1 30 rest nesults m==ary 31 Cheetcal and Radiocheetcal 3 1.1 Purpose The principal purposes of this test are to collect information on the chemistry and radicchemistry of the Reactor Coolant and Support Systems, and to determine that the sampling equipment, procedures and analytic techniques are adequate to ensure specifications and process requirements are set.
Specific piirposes of this test include evaluation of fuel performance, evaluations of filter desineralizer operation by direct and indirect methods, confirmation of condenser integrity, demonstration of proper steam separator-dryer operation, measurement ar.d calibration of the off-gas system and calibration of certain process instrumentation, if required. Data for these purposes are secured from a variety of sources:
plant operating records, regular routine coolant anklys.ts, radiochemical measurements of specific nuclides and special chemical tests.
3 1.2 criteria Level 1 Chemical factors defined in the Technical Specifiestions and Fuel Warrant- must be maintained within the limits specified. Wacer quality must be.
known at all times and remain within tne guidelines of the Water Quality Specifications.
l The activity of gaseous and liquid effluents must conform to license limitations.
Level 2 l None 313 nasults Prior to loading fuel, appropriate chemistry data was taken. All data remained within criteria levels except for feedwater conductivity and feedwater copper concentration. These values could have been elevated due to no condenser vacuum, minimum Feedwater System flow, low sample flow rates and l
e
Supplement 9 Pag 3 3 1-2 the normal)v expected higher corrosion product levels dt g initial plant systems operation.
During heatup test condition, these values were within acceptable limits. See Figure 3 1 for specific information on pre-fuel load chemistry data.
Duting the hestup test condition, all chemistry data taken fall within applicable limits except for Control Rod Drive (CRD) dissolved oxygen levels.
Rose levels are expected to decrease during further test conditions with greater steam flow and the steam jet air ejectors in service which will more effectively purge gases from the condenser. Refer to Figure 3 1 for heatup chemistry data.
The Test Condition One data in general remained within acceptance criteria limits. Reactor water chemistry and radiochemistry measurements were made at a time wher, plant conditions were fairly stable.
Reactor pwer was at 175, the turbine was rolling bet with no electrical output load. Analysis of the results showed the coolant to be well within the Technical Specification limits on all parameters.
Radiochemistry analyses of the coolant showed activity levels and isotopes present to be normal for this power level and core exposure. The Dose Equivalent I-131 result was far below the Technical :
Specification limit of 0.2 uCi/ta. In Test Condition One, the stean jet air ejectors were in i service resulting in low condensate, Condensate desineralizer effluent, and CRD dissolved oxygen levels. De high CRD dissolved oxygen level which was of concern during the heatup test condition is no longer considered to be a probles, j It should be noted that Reactor Cor ductivity varied I considerably during the Test Condition One period. ,
Conductivity has, on several occasions, even !
exceeded the Technical Specification values of 1.0 unho/cm for sevecal hours. It was determined that the increase in conductivity was related to placing the Generate" on line and increasing Generator load. One possible explanation was that operation of the Generator was causing the paint that was ,
pt eviously used to coat the internals of ths i
' Moisture Sepkrator Reheater (MSR) and the Main Turbine to be carried into the condenser hotwell, I t'.us causing the increase in Reactor conductivity.
Another contributing factor was felt to be the ,
l Krylon coating that was previously used as a j l
l Supplement 9 i P ge 3 1-3 i
preserrative coating for the turbine blades, which !
was being worn off the blades and into the
, condent.or. Further investigation discounted the krylon coating (due to it's chemical makeup) as a cause of the conductivity increase. This situation seems to be laproving as the plant continues to operate for longer periods at increasing power i levels. gfforts were made during the condenser outage to remove paint from accessible areas in the M5Rs and LP turbine exhausts. Mechanical cleaning by wire brushing and vacuusing was performed on the .
MSR's interior shell surface and hydro-lasing of the t three LP turbine ethaust extensions to the condenser was performed.
Both Condensate Domineraliser afflueni; and Feedwater dissolved oxygen levels at Test Condition One were less than 10 ppb, which are outside of the limits of ,
20 1 02 $ 200 ppb. The 20 ppb minimum oxygen
- concentration has been :ecommended to establish and nr.intain a protective magnetite film on the inner surfaces of the carbon steel piping and equipsont of
- these systems. De probles of low condensate /
feedwater dissolved oxygen has occurred during the j startup of other operating plants. The resolution at that time was to simply continue to monitor these i parameters at higher power levels to see if the
- levels would increase with power. If dissolved
- osygen levels do not increase to greater than 20 ppb by 100$ power, it may become necessary to inject
- weuld first be made as to the corrosivity of the .
water to the carbon steel piping to determine if this is necessary. (
l
) All gaseous and liquid effluent samples obtained during performance of this procedure were within the l license limitations. Various radioactive gaseous ;
effluents were analysed during Test Condition One, t l Grab samples were taken in an attempt to correlate i analysis results with actual sonitor readings. l However, the activity levels being seen at the -
- off-gas and ventilation sample points were still too low to provide meaningful data. Only one noble gas t 1 was detected, at a level which was just above the ;
i minimum dettotion limit. De off-gas monitor !
! readings were also still quite low and variable. l
! Low ot*f-gas activity values are normal and expected ;
! at this power level and core exposure. [
i l
1 l :
1
i
- Supplement 9 l Page 3 1-4 j l
A measurement of radiolytic gas in steam was also made at Test Condition One. Analysis results were
. below the 0.06 cfm/NWt limit. Radiolytic gas is the !
hydrogen and oxygen formed in the reactor by l radiation induced breakdowr. of water molecules. :
Values higher than 0.06 cra/NWt could exceed the I capacity of the off-gas system recombiners. !
l
- See Figure 3 1 for more detail regarding the >
chemistry data taken during Test Condition One. )
i he Test Condition Bree data, in general, remained f within acceptance criteria limits and satisfied [
Technical Specification requirements. t i
Reactor unter chemistry and radiocheatstry i measurements were made at a time when plant i conditions had been fairly st,able for 48 ho9rs. :
] During this time period, the plant power level was i held between 43 and 45 percent. Some of the i chemistry results, while still acceptable, indicated ;
problems with the primary system and especially with [
- the reactor coolant cheatstry. Approrisately three f hours prior to taking samples for this test, [
Condensate Filcer Desineraliser (CFD) "B" was l removed from service and CFD "F" was placed into '
service. Reactor water conductivity spiked, from t
- 0.58 us/en up to 0.82 us/ca. At the same time, '
j sulfate levels increased in the coolant and the pH dropped. Since all of this occurred in the same J time period, the conclusion was made that there was a resin intru11on and that the CFDs were the source of the resin. Numerous other chemistry escursions have occurred which support this conclusion.
Following those occurrences, progress was made in reducing and eliminating the source of the resta intrusions. he procedure for precoating the CFDs was changed to allow for a fiber underlay on the vessel septa. This inert underlay was used, on an interia basis, to reduce the amount of pawdered resin which was escaping. Since that time, elements (septa) of a new design have been installed for each of the seven vessels, he new design septa utilises a porous metal nesbrane which has a very small pore size, when compared to the old design wire scesen aesh elements and the precoating procedure has been changed to eliminate the use of the inert fiber underlay. No further evidence of resin intrusion has been noted since the new septa have been installed and as a result, reactor coolant chemistry has shown significant improvement.
Supplement 9 Page 3 1-5 The higher than desired levels of sulfate in the reactor vessel were utilized to complete a reactor water cleanup (RWCU) test which could not be accomplished in TC1. This test was to determine the chloride renoval rate of the desineralizers. A test procedure revision was made to allow other anions to be used as well as chloride, as they would have similar RWCU renoval rates. The RWCU successfully demonstrated a removal capability of greater than .
90% for sulfates. !
Condensate and feedwater chemistry were also examined. All values obtained, with the exception of dissolved oxygen, wara within the water quality ,
spectication4 limits. Again, however, some of the i' results reflected the problems which were occurring in the primary system. Condensate conductivity was higher than would be normal, and this may have been attributable to carry-over pf volatile resin i breakdown products in the eteam. Feedwater !
conductivity values were also somewhat higher than normal, and again this may have been partially the result of resin breakdown. Resin escaping from the condensate filter desineralizers ould be exposed to high temperatures in the feedwater system, which can ,
begin the process of degradation. The insoluble l I
1ron and total metals found in the condensate, condensate desineralizer effluent, feedwater and reactor water were within the specification limits !
and at levels expected for a plant startup.
i D e two exceptions noted during Test Condition Three testing are identical to two from Test Condition One. All are for low dissolved oxygen (< 10 ppb) in ,
the condensate desineralizer effluent (CDE) and in the final feedwater (FFW). A minimun level of .
I dissolved oxygen (20 > 02 < 200 ppb) is desired in '
the feedwater system Eo promote and maintain a passive cerrosion layer on the pipe walls, 1,ow [
1evels of dissolved oxygen can lead to excessive corrosion and higher corrosion products in the '
feedwater samples. Current corrosion product levels cannot yet be conclusively attributed to the low (
dissolved oxygen, but if the dissolved oxygen level does not increase with increases in power, it may be necessary to inject oxygen into the feedwater system, hese parameters of dissolved oxygen and ,
corrosion products will continue to be monitored closely in future test conditions. ,
l a
l 1
Supplement 9 Page 3 1-6 All gaseous and liquid effluent samples obtained during the performance of this procedure were within the license limitations.
Various radioactive gaseous effluents were analyzed during Tc3 Gre5 samples were taken in an attempt to correlate analysis results with actual monitor rcadings. However, the activity levels seen at the off-gas and ventilation points are still too low to provide meaningful data. D e sua of six noble gasses is plotted against the off-gas monitor readings, but the plot has little meaning since present off-gas activity is too low to affect the ,
moniter. However, the activ!ty is sufficient to perform an analysis or the off-gas radionuclides and reactor water lodines. By normalizing the nuclide activities with respect to release rate, fission yield, and half-life, and then plotting the data, it was determined that the plant has a "recoil" pattern of release. Such a pattern indicates that there 15 no failed fuel.
A seasurement of radiolytic gas in steam was made.
Analysis results were below the 0.06 cfm/MWt limit.
Radiolytic gas is the hydrogen and oxygen produced in the reactor by radiation induced breakdown of water solecules. It is a normal expected process, '
but values higher than the limit could cause the capacity of the off-gas system recosbiners to be exceeded.
See Figure 3 1 for more detail regarding the '
chemistry data taken during Test Condition Three.
Also note that identifying marks have been added to several data points in Test Conditions One and Thrce to note that sampl6 dates are other than that of the main column heading. Reactor power conditions were, however, approximately the same as during the balance of sampling.
The Test Condition Five chemistry data was actually recorded during Test Conditior.Three as the plant achieved the required power level for the test (65-80% CTP). The requirements for sampling are based upon reactor power level only. In general, -
the data remained within acceptance criteria limits.
Supplement 9 Page 3 1-7 Reactor water chemistry and radiochemistry measurements were made immediately after raising '
power to greater than 65 percent. The test results showed that the reactor coolant chemistry was satisfactory at the time the samples were taken.
Later data points taken after forward pumped drains (FPD) had been placed in service showed an increase in conductivity to about 0 38 us/cm and a rise in sulfate levels to about 80 ppb. nose relatively high values can be attributed to the FPD water being sent directly to the reactor. This was essentially <
the firJt time that the FPD had been utilized on a continuous basis.
ne conductivity and sulfate values peaked at the i levels previously mentioned and then began a slow decline. The source of the sulfate contamination is from the FPD system piping and equipment upstress of it. It is known that the internal surfaces of such of the piping and equipment in the plant had been painted with protective coatings, nis coating and the impurities it contains is released with each increase in power, temperature and system flow. In the present plant condition the FPD water is no longer cycled back to the hotwell and filtered through the condensate filter desineralizers.
Rather, it is pumped, untreated, into the reactor where the impurities are concentrated to the levels observed.
Witn the exception of this first period of FPD in service, the reactor coolant chemistry has been maintained at a very reasonable level as compared to ,
previous test conditions. Pri)r to the startup in October 1987, the plant was in a several month outage. In that outage, all of the condensate j
. filter desineralizer elements were replaced with a l
new design element (septa), as previously discussed. De new elements have prevented the resin intrusions which have occurred in the past, l and which caused reactor coolant chemistry [
problems. Even with power operations up to 70 '
percent, conductivity was maintained below 0.15 uS/cm prior to the placing in service of FPD. i Condensato desineralizer influent, effluent, feedwater, and FPD chemistry were also examined as part of the test. All values obtained, with the exception of dissolved oxygen, were within the water quality specifications limits. Condensate .
desineralizer effluent water was l
l
Sapplement 9 Page 3 1-8 excellent, with conductivity equal to that of theoretically pure water. FPD water and feedwater conductivities, although within specifications, were higher than that required to maintain the reactor coolant conductivity below the water chemistry guidelines value of 0 3 us/ce.
De insoluble iron and total metals found in the condensate, condensate desineralizer effluent, feedwater, and reactor water were within the specification limits and at levels expected for a plant startup. FPD insoluble iron was elevated, as would be expected for the initial operation of a system.
All gaseous and liquid effluent samples obtained during performance of this procedure were within the license limitations.
Various radioactive gaseous effluents were analyzed during the test. Grab samples were taken in an attempt to correlate analysis results with actual monitor readings but the activity levels being seen at the ventilation points are still too low to provide meaningful data.
A sample of the offgas steam was taken, after the two minute delay pipe but prior to any further treatment. This sample was analyzed for its noble gas activity. We nuclide activities were normalized with respect to release rate, fission yield, and half-life, and then plotted. From this plot, it us determined that the plant has a recoil pattern of offgas release. Such a pattern indicates that there is no failed fuel.
Three test exceptions were taken during the tasting. The test exceptions are similar to those fra TC-1 and from TC-3 Two are frr low dissolved oxygen in the condensate desineraliser effluent (CDE) and in the final feedwater (FFW). In addition, the third test exception documents the low dissolved oxygen observed in the forward pumped drain (FPD) sample. A minimum level of d17 solved oxygen (> 20 ppb) is desired in all of these systems to pronoIo and maintain a passive corrosion layer on the pipe walls. Low levels of dissolved oxygen can lead to exces::ive corrosion and higher corrosion products levels in the process streas. Based on data obtained to date, current corrosion product levels cannot yet be conclusively attributed to the 4 low dissolved oxygen.
l
Supplement 9 Pag) 3 1-9 Since the completion of the above noted testing, the plant has operated at power levels of up to 955, most notably during the 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Commerical Operation Run in late January of 1988, After the completion of that demonstration run, adjustments to the venting from the number five (5) north and south feedwater heaters were made by plant operations and chemistry personnel to attempt to increase the dissolved oxygen of the forward pumped drains water and, therefore, the final feedwater. This action was not successful in raising the dissolved oxygen in these two streams to acceptable levels, nor has there been any favorable indication that the dissolved oxygen in the CDE w!,11 increase with further increases in power and therefore, a design chan3e has been prepared to add an oxygen injection systen to the CDE if this proves necessary.
The parameters of dissolved oxygen and corrosion i products levels will continue to be monitored closely in Test Condition Six.
See Figure 31 for more detail regarding the chemistry data taken between 665-71% power.
During the Spring 1988 LLRT Outage, major repairs and modifications were made on the MSRs and during that period, while all renovable internals were removed, all accessible interior surfaces with evidence of paint coating were thoroughly wire brushed to remove this material.
This his significantly decreased the rate of sulphate leachate entry into the feedwater stream when FPD are in service.
The remainir.g testing in this section not yet completed is the Test Condition Six steady state data collection and the Reactor Water No Cleanup Test. i i
t 1
t N * .- N N N h N e te O
- v w *
. . O. .
O. O
- E
~
g
, 8 1 1 1 2 8 1 8 , i t i ;
i 3 3 3 _ _.8
_ _ _ 1__3 _3_3_*-
_ _3 _
i i-i i e * .
J , '
> g u
e N
- h@ s t M k * *
'i t
e a
- 4 N 4 W w w w l
. * * *
- M w .
- e e . $. c e M h. e e. . % l.
- e O O v N.
- b. O.
- d O 3 [
> e .
b W N. h $ h b 1
W h
[
) M
- P* [
S @ M N * * ,
% .E > @ s w w w w W
- O e w a >
- e. M =.
s e *
- 4 J e w *
- h p. M O N % i O O
- W* O- 3 %
i 9 * % 9 @ w O,
@ l p e M
. O.
M M N.
- j u u i
- . _ _ _ , , _ _ e., , _ __ _ , _ , _
g 1
- i =
J g h *
- e
, @ @ O O .
- e g c g a w tw e w . e T
- * % at .a N N e w O
e @
e- %e e6 M t 4
- k
- O O -
v - @ - %
- O O v @, 4, M e O-
- e 2 E
- W
- I g [
O 6 - -
! g ;
u .
- __ _ _ _ _ _ - _ _ _ _ .. *_ _ __ _ _ _ e.
C
. . .a 8 8 h
- e 5 i
- 9 e, w
- e O O e g e 4 w
- % se o e e @ w w w tw k M
-l
. @ c O O - - -
f g h O- W- 3 3
- v 9 O M N O' *
% c - . O. w e
i
> > N 9
- i i ,
- 6 w b 1
,
- e e e M =m C e j j % 9 ev e e 4 4 e e 4 m e O =
- b e e % % % % % % O h %
- d .
- O O % C C- 2 * * *
- 3 E w
. .e % E O- O-
- j b e *
& a g *
.l __ _, _ _ _ ,. . * . _ _ _ e .
- i .
b 4 ai e'
- es o M
' m m E E
- or e i * % m% e b i
8 E * * **
j % % .h. * *
- m a * .u * *
- w E 3 w
- 3 =.
4 4 e e c U V 3 w %
- w 9
!
- h a 3 i
- E er e 9 .
p
- 3 w w en . A w . g w e. %
> $ > W e w 4
- w .
4
- h 3 e. M g 4 w a w
- w e j > . e m w M b d L e .
f e C
- 5 w M.
e ** $ en W ** b h C.
- & 4 9 w > p * .
- W
- 1' . 6 * % . e e t .
- b
- b w . W y . W D
- 4 E V b U C w & 3 4 i
- D . O e *
- 4 .*.P.. *. ,
b - C 4 6 I e w C # **
S
- 4 .h
- C %
- w 4 4 9 * .
- s 3 *
{ b m. .
1 W W > W a.
.E.l d . .N .P. *. . i .
j d. D .
D =
y*
. I i - i O b = = =
j > .
d6
- N 4 4,hl thi I
$ b O
- .- _ _ 2 3
_e ___e_ e. . _ _ _ . _ _ . _ _ _
t
+ e M N h , , , - -
o .,
e- - . . . +
O .
eJ V V F., .
k 3 __3 _3 3 __3__3 2 i
s 2 w
8 i u 1 i N $ > k k 4
- h. W . e.
. .a - . O O O .
- O h h O 9 0 9 4 e e # **
O. W W W e h I e M am 8 *m * 'O f f O f 9 M b *
> 9 9 N M. S. e. ==
s k
% Q h s= M S en y N g e
me sus asmi een oss man - em eum ens emi gum suaq em een een em me ama em esa em aus 5 M 8
% % b K (e 3 s O E @
@ ,S * >
.). w - e sk. O e C O+ O 4 4
= g
- N -
O .e
- = % % b 0
- 8 e. % 9 4 @ O N 9 M E E e e- a
> e u y
- v l U
- _ - - - - - _ - _- _ __ _- -- -- _- _- q
- . se m
De @
g e e e * *
, C @ C e =
O S E S
- 0 0 0
- 9 es
- O S M 9 2 % 30 0 O O - O w e
w e
W a
8 -
- = e- O 9 k -
8*
O @ h h
@ 9
- 4
_4 W h n
[ O e e . C. g .D g g
- U t @ M e 4
U ** 4
. E &
I
_- - - - - - - _- - __ - - _- _- . . - D. , w
, , . . . - M
& @ e O O O 9
- e I 3 e h =
W 6
w e
w e
D 8 i
' ** % W D C* O .e e. e. C *
- e e e, O O O O
M M h O 8 9 e *- 4 @ h. u
! 1 %
e e- - e e e e
v *- e D
. Od 6
- m. == e== me == == .- m em em - an- es= - ===== ===e== == o. =< ** == $ 9
, - + e e 9 h C 1% % 4 4 4 4 4 4 4 4 =*
- tan 5 0 4 % % % % % % % % 4 4 0
- O O % 3 3 2 2 3 2 E 3 C S .*
O s s E 4 i ** e 6 e. e. C
& b 8
_- - q= - - - _- - _- _- _- _- _- -- Og
- C 4 .t _t m
.v .
i p 9 m a e an. em m u h e == *
& 4 2
w 2 9 > 4 4 4 h ee W pe W W 8 e3 4 5 m eI e 4 9 w %
e & 3 % a % 3 % 3 %
- w a u o g g = * *
- U W C b 4 6
- 4 m b a Q U Q U $ $ b 8 G O e-e h
N w
O w $ w w *w 3
w
) w =
e 4 @ C 9 w * %
I .e 0 e
- > a =
- 4 *e - .
ep e *C
> a w $ C - C e e $p 4 - %
e e C - .e .e o O oO 3 p 3 w - > g 9 *
- * & 4 *a >
- 3 til 5 4 4 h a N 8 h e W W h E e > c 5y w m a a & & 4 4 & * * * % w 80v e e e- e e 4 4 4 =
e e 4 e.
e e W == w ===
e e e a C
- 3 Q 6 & b = 4
> 0 4 e- _ e.-
O C $
- O e- = = a w y 8 w
- 4 e C e Q d Q d Q C Q
- O $ 9
> e- e e. > > b d e- *
- e ==
- e, d d 3 4 ** O C 9 0 Ww *
- G C
- _ _ _& _O _ _O _ _O _ _d _ _e _ _> _
4 g 3,,. ,. , ,
e e- t C 9a
- 6 e
! 4
- I $
8 % 9 $ G
>& 3
- d A e 9 e e I .
I ~- M . ,. e 4 en thi les W I 2 1 - = >
_ E _ _ 1 _ _ _ _ _ _ _ - _ _l fff
i
, < < u e
< h N
. . a.
e .
z s a a a : s s a O **
_y_____ _ __ __ __ __ __ _ _ _ . &
. .s .
-1~ 1 b
e ,
. . b b , . .,
.s ah. . . 8 O. O, e W
- s. O.v
- h. . h. a .. . . O. .
8 =. =
@ $ e v O O O e
- e. e >
b h h =
. e. c. 3
%
- N O 9 Y h e 0 n **
b
( e% 0 E
4 4 4 e
@ S e
- w .c , as s Q c O O. O. e. >.
a I, , 4 O. N 3 3 3 w
% g g c w w s v C v O N w h @ %
e e *
- 4 W 3 w
== .- _
F P*
(W @
t g .
, C @ O .. e en e =
- a 0 e w s C O O e n G e e.
g &
- 0
- s d
A O O O w W
e W
s W
6 O.
O
- - h.
- 3 M 0 G - *
- 4 e O y O O 3 wM
( >= %
O h.
h 9 N 4 0 9 9 e * @ t i* f* M == 4 v &
_ _ _ _ _ _ _ __ __ _ __ _ _ _ M.
& + O p. e 6.' N A 3 e 4 O O h
% se o w i O. e C. s e e e .
- e o O O .
w w w s s *
% e
- N '
p.
e -
e >
. g
_ l_
l 3
l
- , .s 9 e C l
3 t % 4 e e 4 g e e e - E
- w e e= e s % s s s s s %
e .e e 0 0 % 3 3 3 2 3 3
- I C 4
O s % 3 g I 9 e
w .. _ _ e . .
b &4.3 @ ea h
s.= Q s %
a y e e e g s e n a w w w w a b eN N 3 C w % w w & *% % e 5 % e s % u- d C O @ t
. u e ae e 6 - O e a u 3 U $ U p w m. C a a k
= - 3 w 3 - 3 h e 4 e o e e a w a
w aw C
$ 8@
w s a w
= @ *. t
> 9
- a m a > a. 6 w e > c e C y > g $0 .O
= = D h & *
- q = 4 % % 4 w
4 w &
e C = e eN es N - E eh 4
- 3 w a $
C D e
8 O h eh h $
9 4
-G % % e G S > 0 & 9 & & $ & * %
w
- b $
w 3 4 C O 4 4 O C y
.i e. .a .4w. - e- .
3 6 4 = a e K - - O =
.. . . 1, .'.l . v .
W Q 8 C D Q e.t O O ,S
- 9 6 e Q 4 w p e. 3 = *
- 6 e 4
- 4 9 4 4 4 9
.C.
- Q Q C
_ _a _ _a _ _0 _ _v._ _U _ _C _ _>0 gw e 8
.h m >
E 6 e I d
=
K e d
> S 0 e
> $ 6 . . . ., .
e .e 3 . C h
- e e <
g - e s
% b G b d W I e a W e. *
- g q) w w , oo i
_ ._ 2 2 1,
{
1 u w e r e n u b
- n
. . . 8 . . . .
. . ~ .. . . .
- s s e. e. O: s a a e a a e a 3 3 - - -
i 3 -3 3 -3 8
.C i
g
- 8 g
}
5
~ ,
@ s ,
.I .%
, . . ~
> s . . . . . .
i .
O
~ , , O,
. :. . O. O. .
. ~ ,. , ,
j ,
g g g e O O. O O v O O h O. v -
g ;
l
. - )
u h e.
- vl ,
i
= = = = = -u == === .- e- .- .= u. .- e- == - == -
0 n b a w C e O e * * .
.a i
6 % e f s a h 9 . p *
- n. O. 6
. *
- W e @ 4
- e. - . , O. N . O. O. O. e. O. N , K
' S % 9 9 @ O O O O v O O = 0 v h 8 3'
> D @ =
C e e e e e
. e e e e e 4
- a V $ r J - - - - - _ - _ _ - - _ - - _ _ _- - _ . I t
. y
- la 2 e ~ .
e +
. , e . .
1 .
- n .
g .. % e p.
n - 8 8 0 u. , 8 -
N O.
9 h O O - . - -
> 8* *
- 9 * @ O O v C C h O v n K
- [ W 9 .i',
> D % @
- l g
- O
, C D e e o e e e l
$ = $ . O P g *
- ee n ?
Y E &
_ - - - - - - _ _ _ - - - - - _ -- _ _ a w
- v ;
Q c 4 ,
e C m M 4 *
- 2 D !
% $ .E 4 Q e 4 O M e N
2
- a 9
- 9 @ @ O O % - - %
> 8 9
- v 9 2
- O O E O = N O v n w I %
e =
e C
- l e e 3 .-
. e b [
_ - _ - - - - - - - _ - - - - - - = = _ 4 > $ 0
- # . 4 .
e e D W 4 C ID
. 4 e e 4
e e -
g e
9 -
O e 6 4
~ W t
i O 2 O 9
- O v I 4 * *
.
- O O % 3 e 3 J e .
l 0 w % 3 6 - C p
b =
% * . 9 l e,
gfG,j
= eee C
ka t
a . E % % 6 m 3 > N n* I a 5 a a 9 N N ;
f W D .e %%
5 a % w a u e* O $ $
m . O E
- OA t
. a w 4 w e w - =* a
- e 4 w h
a D J 4* 4. 4I %b. I e w C (
m . w e C.
- t = 8 @ h D i
- C 6 a y 6 d . e e e C W > e e e. .m u - .a a e. 3 e t % %= 4 l e C b
- 4
- E w C l
= .e & g e. e 3 u 8 W > w 4 9 9 .Z N %%
N MW I
== K I4Q u >
.e 3
= *
.OC v' C
t 3 t * %
9
.t 3
. O @ a
. . O =
(
.e ._4 1 . s .
. . &,.% . 3 w -.
s .
-
- 4 9 9 i
. C e C . O K C l, *. . D D C 8
_e _= _. _ .O
. o.
_e >_ _v _ _v .= .
6
- e. . u o e3 o. ,
.!* t
. ~ , e O '
w w 2 N I
.e. -. .- .- .- -. -. 2 2 ,
n
~ ~ ~ ,
- . 8 n
e .
t :
O a l a
_ _ _ _, _.. i_ _3_3 __ _i ___3 &
i 8-1 e_
s n. ,- e - .s - -
2 . . 2 .
,;g g
g g - =
W N. .
h
....s e. .
~
.- % 9 @ E 3 3 2 I
-3 .
4 u
- 8 , < <
a .
3 %
k A C O 8 **
- ea 9 3 2 3 2 3 W@
i . >lv %
e e
E &
3 .
. == == w
% .E . . .
- e e D C % % % % %
- 9 9 2 2 I E 2 Y $
- e .
. O % C*
3 % 4 4 4 4 4 b o e % % % % %
- O O % 3 3 3 3 2
. .e % 3 .
k k * %C
. m gi 6 -
E 3 e
. .[
q e u e e gg e
a w I
- I 2
s.
- 4 l
I v -
I k r -
.g:>
i
=
e
$,]
e & e . v 5 = . .e C.
. > - *O w 9 w a* =
8 Q Q @ . @ Q e
4 3 = h 4 8 S
. k. .*y s3 ,: ,.
e = b a 3 - @
d w 4 C J _ _C. _u _
_ _O _& _ _
g 1
l *I
.=
}
4 :.
w b
a . >
j e
>d I
1 , . . .
r k . - ~ <
w %
l W I e ; >
i
== _d .S
r Supplement 9 Pag) 3 2-1 32 Radiation Measurements 3 2.1 Purpose The purpose et this test is to determine the background radiation levels in the plant environs for baseline data and activity build-up during power ascension testing to ensure the protection of plant personnel during plant operation.
3 2.2 criteria Level 1 i
The radiation doses of plant origin and the occupancy times of personnel in radiation zones shall be controlled consistent with the guidelines '
of the standards for protection against radiation outlined in 10CFR20, "Standards for Protection Against Radiation", and NRC General Design Criteria.
~
Level _2 None 323 Results Radiation measurements were taken in the form of process and area radiation monitor data and site surveys. To date, all data taken has been acceptable and personnel radiation protection has been provided in full compliance with the criteria.
Sea Figures 3 2-1 through 3 2-3 for applicable sonitor and survey readings. These Figures reflect the results of this test for all the test conditions for which this data was required.
I l
l l
l
[
Supplement 9 Pag] 3 2-2 FIGURE 3 2-1 (Page 1 of 5)
Area Radiation Monitor Sensor Locations Diannel No. incation (Col.) Floor-81dg.
1 (F-10) 2nd Fir. Reae. Bldg. (RB) Pers. Air Lock 2 (B-9) 1st Fir. RB Equip. Air Lock 3 (J-13) 2nd Fir. Aux. Bldg. (AB) Access Control 4 (G-10) 2nd Fir. AB Change Area Control 5 (B-13) 3rd Fir. RB CRD Storage and Maintenance Area 6 (G-13) 3rd Fir. AB Main Control Room (CR) 7 (F-9) Sub Base. RB 3.E. Corner 8 (B-10) Sub Base. RB 3.W. Corner 9 (B-15) Sub Base. RB N.W. Corner 10 (G-17) Sub Base. RB N.E. Corner 11 (G-11) Sub Base. RB HPCI Rs.
12 (F-11) 1st Fir. RB Neut. Hon. Eq. Rs.
13 (F-10) ist Fir. RB Neut. Hon. Control Panel.
14 (A-11) Sub Base. RB Supp. Pool 15 (F-15) 5th Fir. RB Fuel Stor. Pool 16 (F-15) 4th Fir. RB New Fuel Vault 17 (F-12) 5th Fir. RB Refuel Area Near Reactor 18 (F-13) 5th Fir. RB Refuel Area Near Reactor (High Range) 19 (L-12) 3rd Fir. Turbine Bldg. (TB) Turbine Inlet End 20 (R-10) Base. TB Susp 21 (N-7) 2nd Fir. TB Main Cond. Area :
22 (J-4) 1st Fir. TB Decon. Area 23 (M-17) ist Fir. Rad. Waste Bldg. (RWB) Control Rs.
24 (N-17) Base. RWB Equip. Drain S. Pump 25 (P-16) Base. RWB Floor Drain S. Pump 26 (R-17) ist Fir. RWB Drum Conveyor Aisle Operating Area 27 Spare >
28 (G-11) 4th Fir. AB Vent. Equip. Rs.
29 (B-15) 4th Fir. RB Change Rd.
30 (H-12) RB Basement Air Lock 31 (B-12) ist Fir. RB Drywell Air Lock Labyrinth ,
32 (G-13) 1st Fir. AB Near Blowout Pnl. :
33 (C-9) ist Fir. RB South Af.r Lock t 34 (N-2) 2nd Fir. TB Near Off Cas Equip. !
6 35 (R-2) ist Fir. TB Near S.J.A.E. Area 36 (K-1) 1st Fir. TB S.W. Corner 37 (M-2) 3rd Fir. TB South End 38 (R-14) Base. RWB Scrap Cement Recovery 39 (L-13) 1st Fir. RWB H.P. Lab 40 (P-16) 1st Fir. RWB Receiving Area 41 (5-17) ist Fir. RWB Balling Roca 42 (N-16) 1st Fir. RWB Filter Desin. Area 43 (S-17) Mezz. RVB Washdown Area 44 (S-12) 1st Fir. Service Bldg. (SB) Mach. Shop. {
l f
Supplement 9 :
Page 3 2-3 FIGURE 3 2-1 (Page 2 of 5)
Area Radiation Monitor Sensor Loostions i Qiannel No. IAcation (Col.) Floor-Bldg.
'45 1st Fir. Inside Drywell i
'46 1st Fir. On Site 5tg. Bldg. Control Aoos l
- 47 1st Fir. On Site 5tg. 31ds. Compactor Roon !
'48 1st Fir. On Site Stg. Bldg. Truck Unioading Station !
l
'The remote indicator is located on Process Radiation Monitor Psnel H11-P884 (Relay Roon),
i I
P l
f l
r i
i
'M1
-.m .e e m .e om - D. g - e -e - W e. e I, W Ob e
C e w 9
% C e w .C m g N.
e w # 4 N O M 9 O e C @ M @
e-y y
@ O n o n a
O.
O. O.
.. O.
C C O M. (. c O M. C. O. O. O.
hM C @ O n e O O. O O O O - *
- O O h O O O O O 6 0 @
- O % N O V h 4 b
M %
b C # h
- 6
- n m a @ @ @ n e h h t m 9
- k h,
o @
O. O. O. O. O. O.
M. h.
== 0
=. o 0 C. O. O. O.
p , e n . .
e= h a h a o O O O O O O O O O. O O O O O O O
- p. . h c e -
C k
- b %
W h m ei - - a h i e - - - e.m .e1 - e., .m. -
h C >
0 E e @ e e N E M N 9 T N e @ @ m @
- W e* O %
p , e er o e C. O. O. O. O. O. O. O. O. O. O. O. O. O. O. O. O.
t
- 4 9 - % 1 O O O O O C O O O O O O O @ O O O n- $ f n : N S g Y
E s' 6
g
=-
a g p e
- 4 e: e o o k.
8 8
- en e h E e W 9 h h h h h O 9 @ T @ e N. o e e O O = 0 0 0 M C W M g .e p e g C C. O. O. N. . M. N. m.. o O. O. O. O. .- 6 m a D. T *. 9 - C O O O O O O C O O O O - O @ O O O
- 4 8
- b ""
W C * ( s
- 9
, y O
O.
S O *%
- 3 4
g
- 8 E =
- g, .
' 1 e < . s * .
g 3 E c h N h E 9 9 h #
- # E % C O O O O O h 9 M N e O O O- 8 w]* #
q @ f C O F - - + - - - - - - - -
L g 9 e O g
0 0 O C O O C O C O C~ e O 9 O C h I
9 8
- C g I g - - - -
( -
h 8 8 3
- C
- D 4 e o e C e e s % . . . . . . h g m N e , , * *
- 4
- C O O @ - - - - - - - - - - - ~ -
D * >
- O - 0 0 0 0 C O O O O O O g O- m O O' O 4 0 0 0 % v v v v v v v v v v D A
> 9 h 4
_v
=
i b 4 e a + m m N a M e N g 2 6 3 t g a h h w w 6 m e o o o % O O O O O O N N M N e O O O O C 6 # C C C S - - - - - - * * - - - -
es
- O O O O O O* O O O O O' e o m O O O e e 4 6 C. s e e- *
& y = 4 S
- b bi 4
- - - - - - - - - - - - - - e
= b -T t 4 C e E g e- 6 e
. ] ge.
. O .
I
- r.e m. 4 ( @
h.
- O e e
h n g c @ h a p O. . . .
y p. ..
6 8 s. N M
- 4 T * \
y &.
s J C as we W W 4 V a 3 C1 l' * * *
- i O O O
. , - Wi . - - - -- -- 2 2 2 m
t * @
C s C i C %
- M 9 N.
e e pf E
- m m m A 9 5 C # 9 h e C C- 9 E M e
- p e 0
O h
= C C
O
-' O e -
C C
C - - C. C. D. - .e.
(o @ C e
- m r* M C C O O C C C C C O C @ C C @ t
- a 3 p.
e m %
k C $ =
N N N 9 h 0
- h & .
- b @ C C O C. h.
- e C. C. C. C. O. C. h.
a , a m. o.
p p , o C. o C. . . p e
- h e rc e m
- h h o O O C C C C C e 9 m o O
> t -
e e c
- C C h e 6 k > %
W h.
e
- e ed 4 d a aa a a .m. e p. a .m.4 a e
.- a e. d a e.d h
C k 0
- e e C-m M C m N e N
C e
C O O C C. C h h
- W ( 0 % C C m. C.
P.
- e h C C C C% - - -
C - - - -
t
- 9 9 - % N C C 4 C O O- C C C C C @- * # C C
> c m S
Y c
e i.
a
- a. g *9 O a # 4 - @
, C p % 8 8
h
- 0 C 4
- h w h.
e .d v M M @ h C 9 + o W M e e a O O o g
. O M M O. C. M. N.
h @
C O. C. m. 9 C. C. .
- s 9 ee ge C 0
- c. .
C m C O O C C C C C oC c C C D 3
- W
. *
- h 9
- dC 4 - 6 g {O
- c%
O
- -C 8 - S
.I
,1 y e
- e e g
- E =
g - - - _ - _ _ - . ., . _4 g _ _ _ _ _ _ - -
G f 8 0 m m
- c 9 N N
- g 3 E
- . 38 E % C 9 m C M f'" h .' O O C N O C S 7 8 m C C- C' - - - 4 0
- C D R
pe g f - - -
C C ( C C
- v C
- e D C m C C C C C C C C g g 9 %
g < -~ . . 5
@ C e 3 4 *
= c C 0 4 * % C 9 m *
- m M N * .
- h * * *
- 4
- C C D @ C - - - - - -
C - - D *
- a C
- m C C N C C C- C C O C C C 9 O C 4 9 4 C % y = v e C. = v 4 C D
- 9 h 4 l
v - - - - e e p
- 3
.C
- T E e 4 e t C 9 a g e h e 4 m N @
- w w 3 t C m C C e mm # C C C h C C C w e C C o C
& C 9
- . - - - - - C - - - - - -
t 0 $
e w O
C -
e C C h C C C C C C C C C C c o C e e e 4 * .
t *
- g
- 4 e w v & W 4
<- . <..I .- .- - - - - - ..- - t * *
- b D C C 4 C 4 0 g 9
- w i e e t 9 @ *
- l Z; O =
h m 9 e c h c e C - N m 9 > 0 e
- a. e C
- h h h h h h h h h h m e m m e e n t- es gi >
- h m i
e
- L 5 }.
- w w w e e B e i 6 U C l C 0 C
> S li S
-. - VI , . - - ----- - -- g g 2
0 @
= * @
C e C e o %
a e e. # E m. e 9 O N c m m 9 9 N N.
e- t O <
O*
O = O O O O O O O V t- O O n y y O. k O - * - * - O -
- C C O e N. O O O O' O O O* O O' O' O O 9 9 =Q e o @ e @ % 3 v h 4 4
e m %
A C *
- g N g N N m N N o .- W s e m N or m O @ e 9 O O e =,= b, 8 % , O. O. O. O. O. O. O. O. O. O* O - * -
- e. N e N a O O O O O O O O O O O O M @
> V -
O @ O C C N
- O b k %
U N e
to C A o a e m m m N ( m m ( g m w e- # @ O % 0 0 0 0 0 0 0 0 0 C e 4 O O.
O e E N O E - - ~ - - -
C -
e =- 9 4 - % 0 0 0 C- O O- O O O O O O- M M M @ m 0 U 9 9 c _ _ _ _ _ .. _ _ _ _ _ _ _ _ _ _ _ _ _ _
s
- 1
- e. = G a g @
L = f G 6 i % :
.e C ge %
A 0 N e-0 O '9 w - N & = .n m e N N c m 4 m 4 m + 0 e *
- O O a O O O O O O- O.
O e @ w m e= ne O @ C - - O. O. - - - -
O *
- 6 @
r > t 9 - 0 0 0 0 0 0 0 C O O O O O- N N 8 3
- W g C b.- c %
- 4 - a e
. O O * -O # "* 3 3 e e d
- 0 g
. v *
- 5 =a g
- _ - _ _ _ _ - - _ _ = _ _ _
gg, .
] Q # a Q O g 3 E m m N N e m e g 9 N *
- # a % O o o o O OO O O O e 9 m 9 3 8 g 8 ( O O m - - - - - - - - -
O - - b 0
- g e = C - 0 0 0 0 C C C O O- O O O O M e s -
g 1 9 %
- C e .m. - - - - - .- -- - - - _ - - -
a e= 3
- C
@ p 6 e
.- c C 0 C e w % s- , e s. e e w g *
- 9 e o c o 4 .. . . . . =. . . N. . n= >
e 0 0 0 C O O. O O. O O O O O m 9 e 0 a
e O % v v v v v v v v v v 6
> - T k 4
- - - - - - - _ .. - - - - - - - e a e
- L 3
- C-
= O
= b 4 4 e f C 9
) e a e m e N N e m m c g m N
- w 6 m o O O o % C C O O O O O O O O 6^ C O O 9 - - - - - - - -
O -
6
- C e .4 0 - 0' 0 C O O O O' O O O C O M P 8 4 0 6 - % e * *
& v e- 6 4
- G W
- == . - e - = == .- 6 =W
- b 1] V4 b b b t
- 6
- e D bA =
.e g -
b h 6 C m P M mM 9 9 9 9 9 9 4 9 9 0 =
y > .
I $
- N m e- & 4 T =
- > M C 4 -
W W ed a v 3 3 O di e * *
- VI O O O
- ---- ==------ - . - - - -- 2 2 2
E N h h h @
F g a 6 e 4 w
- W w % w w w w *
- h
- - @ @ 0 m C C C C 1: e a *
- h C =
C. C.
i C C C c C C Q O C ( ) -t h e e C C m a 6 C -
( t. w C m
C e
+
- p. C @ * @ h 4 G e N @* m . @. . - ,
C * @ @ @ e 2
=
0 9 U %.
P
- - - - - - ---- -- -
- q h 9 e&
K
- m s e C C # 6 C e 9 . to N r, n *
- P. w t @ w w w w e 8 e
- i rm. = i C
- e h e o.
P
- w C C = c
- C C *
@ PS c O. O. C -
@ c P. P. @ C C C C C 0
- 4 9 @ @ @ N - . . ,
c N ( 0 C C.
p h.
p % r. g g g p. W U h e 9 g
C .* ** h h h 6 C w e w w w w w 4
- p ( O %
- e A C E @
- m *
- O C 6
- 4 % e s C. C.
- C. C. C. 9 E C E c + c
- 3
'" T ti" m N e p. g to * - -
m* (
g g C m' T m 9
=
C C M 3 w
= -- - - - _ - ~- ,- ---- -- .. ----- E G J %
- g h Q
D 4 C o w h
.=
e c h o = a C 4 N@
- * *
- C C a a C C
- d * # 9 4 - w c @ C C C 6
i N .
4
- h ( -
- N - g g N 2 m @ - m N (
- p. t
- s O + 0 C C o' m 0
- w *=
C e' a w 3 .U - - - ----- -,. -- - -- - ~ __
e
- a $
- w e q 3
- #
- D 4 8 C, x > v. L e
- 3 6 s O w*
C h C C
C C C 6
D u
)' e
==
m -
% * # C t* -
N e* I s e c C C
- m M C h 9 P* C - *
- N 9 3 6 e 9 %
g g e. e.
p 9 e h
---
- Gt e - - - - - --- -.-- -- -- -- .
- C 9 3
- V y ( C 5 E = c E 6 w w C C 4
- w C e w a
- C C C e e
& 9 e C O o 4 g N N ( F Q e c - Pe m ( N N e
- h *
- 4 -O O e C w
> e 9 6 "C f v
- - - - _---------- -. ---- - - -- 0 4 e e w 6 e' e a c h 9 4 C C C 9Dv w
. w @
C 4 C-e e s - .
C C 3 t C C C r m m C r= m m * *
- C ei t= ( 4 w e C e 4 6 8 o - %
- o e .a v -
- b * % 4
& 4 - I L
- - - ,------ .-----------,- --- -
- e a m m m - e ~ c e e 4 6 e6 6 6 6 6 6 e *
- g & L m m =-. 2 L L t - - e e C e w * * -
is 6
i s s
e s 6
e c,
y
- c. c.
s s
$ s y e E. g -E. *e e e t C3 C 6 *w w w e> DE 9s es w w.I+
- s-w 0 .o* - - - -
.C - .C -.C - .C -.C. - w -e C
e e
e
.4 .J J .4 a B
+- -CC-
= * -e - ,
- t t t t t t -
C m o e e q r -. t . C 6 4m ee 4m 4m S .* - 6 e C
- e ele c 4 mwm $m ea 6e 6C DU 4 C B el e >m C 4 O 3 a T 3( d @
- f* *MieP *f E C 2
- V O
C (
6 O
e9 a e4 C
0 f*
C,5 Q =0
&*ies a O s% C A O A C e q" lw v% 4 4 4 JI MC ordE3K4K Os.' gO. *CC
- W > aa ea 96 a ea =a a Ca Ca C alC a a- s s a - t J e J4 i e Ba * * *g.* *gwa l* * ==i*=i =' *5 t
.g- si
- 4 wi J .i e we ge e 3*I 3-3Je=
- & 4 y* i ic i i-
>- w w p e 4 4 . e. *e .s- t= 3 s- s) - ia - m- D- * ** *
$ V 3 3 0
- O + O'*
- O 4 O;A Q C'SwasC -44O.lc wy Ol 06 O
- O1* O*O
)O w -a .O O10wO- O C
.O . . .o.----.-
- ------- ~..--- .- J 2 e
O E O C e C O 31 8 4 % C O en @
w w & @ O n C C C e s w - N O e- C C C C C e 1i
- e. .- @ O e e O @ C C O C C C C @. I es.
> . e - e N e
4 9 - @ N ( 4 - C C i n
% e S
.u - .P. . , . - ---- -
- g h 4 E
9> %
P.
t $
O . e e.
en
- P. n s @ O C e ** e P. O = C C e* N e h a O C C C C e m.
> O h t C b C f f C C C C O. f.e O
- e M @ f* f* b6 4 M C C P P U h e
tu C h O m W C
- * $ f O % O C 8
- 5 N C e O C C g C == @
h ( m e = M h h h o
.e, W
- - - , . - - -.. --- ---- .-. ..i._-----
e @
C
- e p %
b e
(0 oh m h O - C C a
= * - t C C a o C C C C *
(. *e c e
.e- P#
- g 4 3 O O C C C C - I e e -
I n - e O, e e < < e C e ce .
- O P.
O M
- Pe g g I W e - - -------------- . - - e a e
= w Q 3
- ** O e t
e O E > - en L 0 s C C 0 C C e ( 6 m w e a9
= m
%> e $ e m C C C C C C C C C .
4 e m3 C C = e 9 h es t. 9 4 9 C C e*
8 9 4 4 % *
{ p h D e e d y - - -.---------------- -- .. -
4
- f y c 7 O e e C b e m % C C C O O h 4 6
- b e O C O f,; C C C O C C C C e e C - P F5 PS f* N 4 4 h C' C 8 e %) * =-
w C.
gI w 4 C
.- - - d >. ---- -. . , . - . . .- --- e e
- w e e e
p ,
w a' C C C C O M.
P. a e e % C f O e C C C C e 3 D C C C o 9 h h =- h 4 & sa C C~ 4 s
m e e - 2 0 - e ea 6
v -
- e a z c
- - .e.
$ e -
e - o e%
gj- gu og g
E. : wt.-
-ga g z\
a
.e e.
e e b'b W V b I hr ,b W g g W
+ - .$. . -w- 6 - - 6 - .- w 6 . .b n
- e .a t o
(
O w es e B. a e e C
') t, e
d= e C C I y e @ b s - p .
er * -O 3m a 4 e 3 e 9 4 (b e e- e 6m Z mJEgC $Ca e D*M{ba I a e a 3e* Q g Tb
( (04 f
- b'> .
- T e.1 C I Z wCl47eJC3e e- **hCC s4- * / i-
@ (4
= 0 *I C $
E E.lw = h g
v O.
r.
Q E. .C
. a .-.
E. . E. .-.E. S . E. -.
c.
. E. - g.C .. . . ..3
- -4 %# - ..
- N
. & a ? . -
i - i-w
- e i =
g w4 .- *t = 1a- 0& * . m Pe i.Th a.it.. a-S-
e > 3 C f. > a - e Z == 3e e w +
e bk 3 3 4 * =C $ e C ++ C *>*
- O = 08 eOd b Ol d C- C f e *-
4 -
6i .C. 3 w eQw,C.dv w,Q -c$ Q.<.c* w *-------- ~.3.5e a - .. -- --
w!=Qw<.- .i O g
1 l
i l!
1!
l' i 1 I l il IiIIIiIliIii1I1lIIIilIIIlIIli 6
8 9 n 8 o % F / 2 2 2 2 t 9 t t 9, 6
- 3 - - - - n -
s t 7 0 1 E t 0 E F e2 e l 6 0 3 . 5 0 m T d 9 9 5 2 4 4 5 5 3 6 7 4 e3 n 1 l e / p e
~ 7 p g i! i l J 1 7
II1 e1I1iII1IlI 1 I L lI '
ua S P 8
3 /
7 n % 1 o 1 7 .
t i 7 3 F 6 2 2 2 2 s t - 7 3 y
1 - - - -
e i 2 - c t 5 F E T d .
9 y 5 7 0 n 0 2 g 1 3 3 7 7 5 6 6 5 o 7 7 /
C 2 t
I ] J l g I II1I1:1II11IlIIiIlII' il 2
n 7 o g 6 t
s t t
n s 5 e 2 0
/
4 -
2 2 2
e t a 4 9 / r t 0 F 5 0 T d 5 3 4 4 e 3 7 3 5 3 n
o a C t i I i 8 1 1 IIlilI1II:IIlI1iI111I1llI1ii1 a 6 D t 8 2 2
% / 2 r n 0 F 7 - - -
)
e o 2 0 1 r t 2 E t t i - O 0 . 5 - 5 2 5 t s t % 4 6 s. 0 f 0 -
n ei 7 5 1 4 4 3 2 5 2 3 3 2 f e T d 1 / o s
e n 0 3 o 1
. 3 n C E e I ] ! I ! 1 I 3 lI1II11IlI . 111h lI1iiI "e R e t t U g t 5 i
G e e p 8 2 2 2 2 w . I P t u F / - - - - r e F (
e d
e t
a 5 O 0
- 3 1 4 4 r
4 d 0
9 E
4 E
4 awi
~
t m
R e 4 4 /
f H 5 7 s o s i s i I ] I l 1 II1IIIIII4 hH11i1i8 I!ilf1iII t h e ut c 5 2 2 O o l 9 t r n e F / - - '
P e s O O
- 6 t t 5 a a
- a p s 0 1 4 4 5 s 4 N w
- s O e 0 / i e l
V 1 4 2
- r b l I I 1 ( l II 1I Ii!III l1IllIlIIiI1 o a t r t e nf 5 2 2 o p l F 2 - - a p m
e 1 r e 0 a a
- a u4 O O '0 9 3 3 a e 4 w h
- e h r F 0 1 o 1 / t o ot
- 1 r t e P al h b i I ! ! l l IilIiilIlIlIlillI1Il{lII!liI1 t s 8
] i i i'i ] ]
r .g r r r a r e s 4 gmh gn h ) h n9n l ) h ) h 1 e a t o/ ti t 6/ e/n) - . 5/ .
5t = t w Sor Sd r n. er h.. t- rasph. Dc n 0r o a. a R =
eoa att
. s. sego(
at( sft 6n e 6*
0It #(
r c i t e l alt t t r t l - - e - - r d o n e o en t t t t d- t t t.s t i n n o St Su n i n t o0 n t i aI i i c e o e o G5 e fa e D e t
i na Op)Oc nh )
v /
t w /
0 .T/ w /
A w /
8
. w)S gt aO a d p n7 u 8 r 9 9 t3 9 l 9 rosD N n m .o4 .r 4 o o) e 0 .
.)6 o no oo se e) 6 0 o) 0 o8 6 o
.) n 3
o e rT na C Y IrC9s a o* p s eI s Lm p8 5 51 % the
- i s
2:
- e 8 . - f .i- - I - - -
.tut t P e d t 91 gt i e vt s e 0 s vl l V' s T e a a t 62 t daz sss eVt t 5 esl eI 1 os t E e C wM st t o uoo vI o r1 uoo bDD 0. s o t T T _ D_ is18(i D 8s( ei( i Ot P( at ( I( ( 0(<O O i l l l l l l1ll1I1lf1IllIljlilI1lIjfl1li N N
b l l glll[lI!9 ll1l H[' ll l
l, l l 6
8 9 n 8 7 7 9 8 7 0 o % F / - - - - - t 1 t i 9 6 0 3 0 0 E E E E E n -
s t 7 0 1 0 0 e7 ei 6 0 3 . 0 0 0 0 0 1 3 6 3 4 m T d 9 1 5 2 3 7 3 5 8 5 6 0 5 3 e3 n 1 l o / 7 8 3 2 2 p e C 7 - p g l l l l i 1 i '
u a 7 S P 8
3 /
7 7 3 n 4 4 3 o %
1 - -
1 7 . - - - E E t i 7 3 F 6 E E E s t - 7 ' 1 0 0 4 3 ei 2 - 7 0 0 1 1 3 2 1 T d $ 2 5 0 0 0 0 0 . . .
n0 2 $ 1 2 5 5 5 6 3 3 2 1 1 o7 7 /
C 2 1
l 1 i l I 1 ' l 2
6 7 8 8 7 n 7 - - - - -
o F 8 E E E E E t i % 5 0 / 0 0 7 5 2 9 7 s t 9 2 0 8 S 0 0 0 3 9 1 0 ei 4 4 1 / O 5 0 5 0 .
T d 5 3 D 2 3 2 3 1 8 3 5 2 n
)
a C i l l l l ll o 1 I t 1 i' a 6 D 1 8 7 7 7 7
% /
r n 0 F 7 - - -
- e o 2
- 3 0 0 E E E E )
t t i - O 0 . 0 0 8 2 5 l s t % 4 6 0 0 0 0 0 5 3 2 5 -
2 a e i 7 5 1 2 2 3 6 2 6 8 3 7 E f s T d 1 / o M n 0 6 8 1 6 4 3 o 1
. 4 n C E o l l l l ! 1 ll l llll l U "e R e i c U g t 5 i G a a p 8 v . I P i u F / r e F (
d t %
- 3 A A A A A A A A A A em
, a a 5 G 0 1 N N N N N N P N N N S i t
R e < 4 /
f H 5 7 s o si s l l I I- l l ' ' ll
- Il t h e ut c 5 O
o i 8 "
r n e F /
A A A A A A A t
a P e s O 0 0 6 A A A p s 0 1 N N N N N
~
N N N N N s O e 0 / i e l
V 1 4 r b
! l I 1 1 l l
l lIll IIl l4 o a t C il
. nl 5 o p
~
. l F 8 mp a 0 / A A A A A A A A A A a 9 0 0 9 N N N N N N N N w~ N e u4 F r 0 1 h r
. o 1 / t o eL < t r t e P al h b l l l 1 l l llI l
,ll ll i Il]l* Il t a
._ } ] ] ) ) 1 t l r r c c c 1 7 t c s i l h ] l h c c 1 c 1 c n c e a v
o e / m o e / / * / / / e / t p g r o p g r t S n S vS wl v S a a c n m c c n m n C e C iC l C C c I a { ( ( a [ e u v u( Do Ou p u i t
- 'r R rI rI R! V [ l
.f ho f d o n n n i 1 - I iI t I - ! . .
o ~A i A A i . g h h S i i w Hvi ) v v Hwl ) g e a a t .l ) .i) .i) d . ld f L e 5 i gDA aDA oD8 gD8 aO8 t h)O ) ) ) n ) 0 A p e 6 r e 6 Sn0 0 t0 t0 O 0 N r
.63 7 6 d r 7 n m e 3 r 1 e e .3 r 1 E1 e 1 n1 n1 t t o e m)8 Ar8 m)8 m)8A r8 r 6 t 6 e6 e6 c 6 C T EhM oM EhM EhM oM omK s M vK vM a m 2
- e - .t - t - t - .f - t u- a.t - - - M - 1 t P e d t .trt ti t . ut . rt t tt cnt m St St .t s T w o a ot nni oi ot nnt aet dht Tt Ti 8ht E E e C m M D C. N O ooD C. S D C. N O ooO et O anO GD GD 5nO T T T C((CMf C((C((CM(RP(RE(l(5(0E( O O l l l l l l Il! lllll lllllIlllll1ll N N e ,l4 11! ,illi l' , i' lI! 3 1)l
iii ,
- j l 1 l l l l lllIll 6
8 7 7 9 .
n e - - 1 t i e s 9 6 8
= /
3 E E t n -
3 s t 7 'c a
1 0 7 e2 et 6 0 . 7 1 m T o 9 3 s 2 T
l e3 n 1 4 p e 3 o /
c 7 p g l I i i i l ll u a 7 S P 8
3 /
7 6 4 n
o s )
1 E
E t i 7 3 r 6.
s t - 7 a 1 2 5 ei 2 - 7 9 3 _
T dn 9 2 5.
o 7 0 2 s 1 3 2 .
7 / ,
c 2 ,
_ t l 1 1 I i f .il 2 .
n 7 7 7 o r 8 .- -
t t s e s 2s *o e
/
8 E E e t 4 a i / 5 5 T o s 3 8 n .
o 1 1 e i I c 1 I I II1tIl t 1 a 6 D
- 8 s r
/
7 8 7 r n 0 -
)
- e o 2 0 1 E E 2 5 t t t - o 0 .
s s e s 4 s 7 4 2 f n et 7 s t 6 6 o o T o 1 /
3 m n o 7 2 .
5
. o t
- E n c e R e 7
e l I I 1 I I IIIIIl c U g .
t t 5 i G a e o 8 v . ! P r e u r r(
l d t s *
/
3 A A em a a s o 0 i N N S i _
t R e < 4 / f w s 7 o s s i s l i I I I l l11IIl t h e u t c s C o i 8 " t P
r ne o o ar /
A A a
- e s 6 ~ s p s c 1 N N i e .
O e o / .
v l
i 4 r b -
o a l I ; a l 1 I1 l1 t c i i _
nl o p 5 m p l
e o r 8
/ A A _ a
- o oo 9 N
- e h r Fu o. o 1 t o o
3
- n -
. et < t e r al
- P h b _
. l I ! I I I lIi1I1 t a ,
l ! l c . c c 9 c sl e a -
/ / t w
~
- t n c s c t io i a a c
e u u t o o t .
v l . t n n -
n 9 i o . 3 s a -
S
=
t -r t so ) t ) O A -
o a 0 s .o O N nt c
n o
i m e n *-
1 6 e.s tem : :
- e t t -
3 2 t P_e o t b .1 sts E E ,
s T n o a : - ns e c u w D ru h. 0 neo T T O O _
r TE(ov(
II l j N N l t iiil .l
,f , -;t :I jjjl4 jj , 114II, }(n 4 i 4 :t'i)1 .j 5 ill16 11id
N
@ E C i g e N
% G
- C M M w .O at E = .e.
w w cs. @ o . N. N N. N. N. N N. N. N. N. N. N. N. N. N. N N. ae e . > 0 N e . g O
> D @ O M
- O O O O O O. O O O O O O O O O O O 3 e C @ * @ % v v v v v v v v v v v v v v v v v H A 0 h U
N E M % C " > s = o > M o . W *
- M N @ N N N, e y N t N e e N. N. N. N. N. N. N. - N. N. N. N. N. N. N.
e- -
@ @ - 0 0 C C O O O O O O O O O O O O O C
> t N @ v v v v v v v v v v v v v v v v v o C O.
P N
- b o %
- V N 3 e e
- - - - - - - - - - - - - - C N +
C N O E e
*8
- af @ o %
N. N. N. N. N. N. N N. N N N. N. N. N N. N. N g p w g N <3 g . . . , e == 4 9 - % C O C C C O C C O O O O O O. C C O O
> t @ M V v v v v v v v v v v v v v v v v C u 8 0
- U O M - - - - - - - - - - - - - - - - .C .
C L 4 3 a g e @ e O M N "f e g C $ % b 4 i M Ol ** - C 0 N & N
= N N N N N N N N N N W N N N N N N = W N. e. >h g
e
- e = #
N
- C o C @
e - O O V c v
~
O v O v
+
O v O O C v v v O O v O v O O< C v v v C C C v v
- @ E
- O 0 6 M C W a
- > V v v
(= ** C O C
> e ac Q e fl ""
v e - e E
- o O me 4 6 4 8
V a @ cw 0 3 e - N, N, N. N. N. N. N.
- o
- p a s N. N. N. N. , N. N. N. N. N. e d n @ O O n O O O G O O C C C C O O O O O O *
- e y 0 - y v v v v v v v v v v v v v v v v o w Z 9 % 6*
@ N C
.-. .- - .- - .- - g 3 6
# # C
= c 2 =
C e E % N N N 3 \ l e e Q d C o o C e O
. N. N. N.
C C O N. N. O C O
. N.
O N. C N. N. N. C O O N. N. N. O C C N. C O d > e 0 e O % v v v v v v v v v v v v v v v v v 0 0
>
- 4 C a g
l .v - - - - . - - - .- .-. g l e e e O f k C \ l = = D e # 4 4 3 V g at N N N N N. N N N N N N N N N N N N w e s e O C o s - - - - - - - - - . O s o C 7 C C O- O O O C C C O O C- C.- C O C- C w e s O e v v v v v v v v v v v v v v v v v C b - - e
& y
- 6 -0 4 =
-. p . .- - - - - - - -
d -g C O V C C 4 e e g 6- C g T G O - N M 9 @ @ N e 6 - C 6 = N M 9 @ @ N e = = = = * * = = = g 4 0 e u >
- N e t
- 4 e D *=
a > a o g w w 4 V 3 3 0 * *
> C C
,e - - - - - --l - - -- - Z Z
s a , , u I , I kd 4
-. --- -.--- --- O M
'wC e=
Wl e e N C %
- h M en .O. af E e -
a @ 0 . N N N N N N N N N N N & 9 g .& &. N O * - * * * * + 0 0
>- W O M N.- O O* O O O O* O* O O O- O 3 4
$ .- @ % v v v v v v v v v v v 0 4 5
'U N
e M % W 5 C .- > E - O M M O
**
- I N N W= N N. N. N. N. N. N. N. N. N. N.
e e N e N . O
- a @ =
0 O O O O O O O O O O C
> O h @ v v v v v v v v v v v 0 6 6 = 6
% ee U 3 N. tj
?
N + C N k e m .O - W W b % N N N N N N N N N N N ' e
- 9 N O @ * * * = 4
' S = 9 9 - % O O O O O= O O O= O* O* O O H D @ M v v v v v v v v v v v C II O O I U 7 e _. - - - .- - - - - - - - - .- . . L
.~, - e U. 3
. e M N
' .a 4 . 6 4 i M C 30 % g es
- O o N E N
.= N .= W N N
- e-y n e e o o N. N.
- N. N. N. N. N. N. N. N. @E MO 1- e- 38 0 @. O O O O O O C O O O O *O F V 4 = v v v v v v v v v v v 0 b WN
,, C N.- @ % >
- 2 O = 3 8 y O.- .= QO
- * .e n
- - - - - - - e E w &
w G 8 Q @ -c e-
- V ' 3 @ N
- O 0 = g g s N. N. N. . N. N. N. N. N. N. N. e d]' 4 @ O @ M C C O O O O O O O O O WW 9 y O = v v v v v v v v v v v 0 =
{ Z T % a= , @ b C
- - .- - - - - - - - - - - e 3 r 6 I @ e C
= @ f e 9 4 % N N N N N N N N N N N 3
- O O O @
- 9
, e O
- O O O O O O' O O O O* O * >
0 e O % v v v v v v v v v v v OO
> - 1 C D l
' + 4
- -- .v .- .-. - - .-. -- - . - t
, 4 e
- g g 7
%C
==
/ U e @ O 4 3 t ta e N N N N N N N N N N N es e IL 4 o O O % * - - O- O O 0 s 0 0 0 C O- O O C O' O O m e .4 - v v v v v v v v v v v C 6 O.- e o I
& v %. 6
- N **
-. -- - .- -. .- - -- .. - - j=
e .e
. . s C
.e* g .
C $ O a N M 4 0 h @@O e = C = N N N N N 's N N N M E 4 0 a a 6 e e 6 = N
** & 4 **
a b B 4 WW . O y 5 O > > <
> C O
.- .- .5 e. - 3 2 t
i a U S a ,.---.-n,. . , - , . , , , , ~ , -- - , , - , - - - , - . . . , . , - , . . ., , , . , , . . . . , , , , , . , . , , - ,,,--,,,,-w, - - - - . , , , , , , . - , , _ , , n,. , , r, , . -
!.i ON i
.g .
E
~mg lv> ,
;) ,s in ,
9(s ll a d Elb as tk "*
- I '
; O j}
- _) i Ei
~
ll il x.. 11 t *il J
- jso w
+c ; 35_ s r1 i _
, 'l$$$, t 3
fj" ,,) ( I_.. '
- I pg .- ;p glp,
,g, a
,f) '
G
.r
; 5
' i.
- - - . * - - - ------- p w -
@ e C 1 e %N*
C % k O M eM w
- s e e a w @ O . @ O N N @ 4 O a O * > O N * *
- O. e. O. e -
O
> % @* O M e O' e O O M N O M *= ." 8 C $ " @ % v v @&
O N U b e M % i. A b i C - 5 &
- i 0 N M O a w
- f 6 h 9 i e e N e N
- N. N. N. N. N. N. N. N. N. N. o 9 * -
@ @ O O O O = 0 C O O O C H D O N @~ v v v v v v v 0 C 6 > 6
, O % w u N 3
- 9
- - - - - - - C N +
C k o E e w
- W @ O % 4 g e g N o g N. N. N. N. @. N. N. N. N.
t
- 4 V - % o O O O N O O O O O O
> D @ M v v v v v v v C N O
V O C 1 - - - - -
- e D L g a @ 4 3 a
, e e o M N
% b 4 I M A *D = w
- N t b
- N N N N N N N N N N = d
% N
- a w e C o a - - - * - - - 4 MO e* W O @ C C- C- C C O O O- O O
- F M T = v v v v v v v v v v 0 6 WT
= @ % >
- cr b o = 3
,0 y O.
9 E
==
ce
& (L e sr a @ f
- E 3 cc - O ,
* # E % 4 N 4 M @ d 4 @ O O M N. M. . . T. T. T. p W
= y C
- O O O O O O O O O O O w 4 % a=
@ C
* - .N - - - - - - p 3
, 6 S @ 0 C Y = e L
- O C D E % 4 4 4 4 4 4 4 4 4 4 3 a e n O O o @ O GW O - O C O O O O O O O O e > '
O e O % v v v v v v v v v v OO
> a y
T C A 4
- - - .i .- w .. - e - D e W e - g
@ e 4 4 e N N N N N N N N N N w e L O O O % + - -
O% o O @ O O O O- O O- O' O O' w i es O = v v v v v v O. Y v v C : 6 - 0 0
& v %.
e b
- 4 w
- .- - - - 4
- u Q
- e D g O t a g e a N M T @ @ N e @ a E 4 I v > a **
t 4
- N p
, & w a > e WW D W 3 O MM O
.e.-==.- -. - . - * . 2 l i
l h r I p i l i
tT t9 a _ J q . . . . r, F , re .a
-z- %., .
r. a E 2
}j - {0
-g
?
3{I
;,, , . .. :. a. :,.u. . : . : w.v.= a.
5 . EO v.. . .w. , . u s$ < F:
. , .y ~k I<.
8 9 [ 'N r il i
~&fi; kk)$
;; 2
. R$.hffIM'iit . 3 h bS t f' d 5 N, ,p".b,ii 'h y./ [r$N ~
I.
\.
Ey
!! '2 lyd Wi-W,,$y,$$,gfli?, ! 6
~~
G- ?,. 9' .
., s.R. ,.e.. det t c ,,
4 . hs . sw c. r l
.pg..w:*/."IIM,?.lfi!>....c. meg g
p j, . .
, , p.
k ,'> I-l %
'N '
/./
c3h fj '
'4 8 D 2l'X -
,..w...,,:.u..mw:+..m...a.-s.n,u..Sals c W((50 c
/
- --- -- - 0 N
'@ e C I p.
e N g
- t' C %
O M M
=
w e W E
- 9 m $ 0 .
N. N. N. N. N. N. N. N. O. e .e $. n o N 4
> y G O M w O O O O O O O O C $
- O % v v v v v v v V W IL 0 h U
N e M % C - M E
- O k M O .
** 0 6,
k 9, g
- N N. N. N. N. N. N. M. N.
e- N. e c . O O O O O C O o C 0 e O @ v v v v v v v 6 N N. P O 3 N. e 9
- - - - - - - - - - - - - C to +
C k O E 5 m e d e o % N N N N N N O N Ww 4 N O O - - * - - 4 e- T T - % C O O* O O* O a C- O e D e M v v v v v v e c O S l U C g . D% C 4 3 a g * * @ e O M N o D G b 4 i M a = C W % j *" w = O O N E N = W N.
- A g e e o n . N. N. N. N. N. N. N. N. 8 E M o 9 - # O W C C O O O O O O
- 4 b > 4 .- y v v v v v v v 0 6 W@
6 b= c % >
- Cr e C a 3 9 y
- e
= Q Og
=
e E a a w g D h C f 3 e - o
.
- W E % N N N N N N N d 4 e Q O M M+ - -
- e d C
e v O
- o C* C* O- c O o- o o =
.e 1 g % v v v aeC o @ > g 3 g - - - - - - - -
6 y 9 C e @ a = e C e 3 C 9 E % 9 T - e o o o @ N. N. N. T. T. T. e i 8 O
- O O O C O O O C e >
0 0 % v v v v v v v v O O
> *= 9 C h y
- 2
.-. - - - m .- g e e e
= Q e @
4 4 3 e & S
- e s e o o o m N. N. N. N. N. N. N. N. O g O O O O O O O O C C O w ea O = v v v v v v v v C 6 - % 0 0 6 *
& y a
. - - . g d =
. . s
- D
'a . g i
t
. e o t
- N T @@b S@ E 9 D .. .
y >
- N I $
4
& 4 D eg WW r > 0 > >
D W 5 Q 0 0
> - .- == 2 2 t
l 1 I l 1 + 1
. - - - -,, -. ,v, ,n,-_ - = . . . .e, _ . , - . ..-n,-... _nn--., ,._n ,- --w-, - ~ , -,
Reactor Building - Basement I Elevation 562'-0" p-=. ecx:r r-.x_. r. .= .=,cg a ~5 r.;;:.. y,. :w.ms=..; " 3 3+ i [ ,.
.. r . . .
i w%4 E .
$g.' j (( '
.c I' 7* _'
-N-
- s. N '
t
' " ., i 4 [
e :
,1 <$ I g
h h ..
,o -
I 6-1 e u
.w .
.- Y -
u cum
! e.'
~
.s m2
~.:
*~
1 E -' h . - .c
- ; GG .
\~.
T G.
/4 Open Haach
. . ".; L $.
.T.
- ip >g
,en .
[n _ c< w , M J ' 6
'*[. _ _
C.::<,5:==wth:.z;
$ ' ~v .
=.c-zu..gzu %~.:dy-%( .
. -n.1.~-&
(FIGURE 3.2-3) Supplement 9 (Page 7 of 32) Page 3.2-18
w -
@ C I m 9 N C % G
- o M e M w e af E a w @ @ O . @ M T T M N N N h N N N C C e N Q@
g . b C N - * - T- + * - - - - - - - - 4* Cl O
> t @* C M - O C C C C C C C C C C C *
- N M a C 3 4 C @ * @ % v v v M CL 0 6 U
N M % at N C - N E
- O 6 M O w
- I N W @= #4 N N N ** 4 4 @ f4 e e N N , N. N. N. N. N. N. N. N. . N.
C, C ' C C C C C C C C O *
> T C N
@=
0 C v C v C v v v v %* */ -v Ov v 0 6 C 6 6 w 0 % W 3 N. ( C b N
- O O C N g C E e
*
- at @ O % "4 4 4 @ @
T. N. e e e N O g T. N. N. N. N. N. N. N. N. N. N. . g g e
- 4 4 - A C C C C C C C C C C C C C C C
- O C O g > t @ M v v v v v v v v v v v v C u g o O Y
I - - - - - - - - - - - , - - .E . m C 6 9 3 a g [ = @
$ 0 M N ,. E 6 C e M g g C yt s O C b % N g W - N 4 N N N N N N N N N N N N N N N N N N ~ W
- e.
. N i OO > 3 *
- s C O = - - - - - - - - - -
C C C C
- G eg 9
- C C C C C C C C C C C C C C C C s > T 3R N 4 - y v v v v v v v v v v v v v v v v v 0 6 WC
( g C - @ % > - g 3 g C C
= D 8 Q -
= 0 O
@ . * ** 1 y - - - - .-. - - - - -- 9 $ 6 CL q - w C @ D +
h @ N N N N N N N 9
- O
,3 p g % . . . N. N. N. N. . . . N. . N. N. N. N. N. d 4 @ C 0 M C C C C C C C C C C C C C C C C C C o. W e
4 v C ~ v v 0
- Q
- I T %
3 @ N C
., .. - - - - - - - - - - - - .- - - - - g 3 6
8 @ 9 C a t e -. s a
% N N < < < , < - < , , < < , ,~ ,- , <
f, 9
- C C O @ - - - - - - - -
C C C C C* C d > Q
- C - C C C C C- C C C C C C C C 6 C. % v v v v v v v v v v v v v v v v v v 3 o v
-gC C4 4 W D D 6
.C
- c 9 @ 4 4 3 e a e N N N N N N N N N N N N N N N N N N w e s er C C 0 % - - - - - - - c - - - - - - -
O 6 O C C C- C C C C C C C C C C C C C C C w C v v v v v v v v v v v v v v C a s C - v s v v 6 % 9 0 6 -
& v -
- - .-. - - - - .,, - - . ~ .- - , , e d +
C O U 0 4
.C g
** D
- 4 .
V C - N M @ @ N- @ @
- m 6 -
@ - N g @ S e @ - * - = * - -
- N N cr 4 C t 0 6 ..
U e. 4 $ N
= a e t =
WW d > ) Q 4 > *- D U 3 3 C C C
> - -- - - -- ----- 2 2 L
O u N
@ e C l e O N C % E w .O se E M
- -9 M
*
- e Ds. 5 o N.
@ O M. O. N. e. ,. M. N. M. N. N. N. N N. % 4 G O ee a@
> D @ O M
- O , O O O O O O O O O O O Z 3 4 C @ * @ % v v v v M &
O b W b-e M % 3R b C N E == 0 6 M O
*
- 4 b b @= @
ee N t N = N. . N. N. N. N. N. N. N. N. N. N. N. O e* @ @
- O O O O O O O O O O O O O @ C
> g6 O b
N b v v v v v v v v v v v v 0 6 0 % w N 3
- U
- e 6
- - - - - - C 0 L O N + V
- C h V a
g O E e 4 -
* * $ @ O % N O N N N N N N N a e o e N O e > N. N. N. N. O U e
e b
- 9 T * % O = 0 C O- O O O O O O O O M O g > D @ M v v v v v v v v v v v v ~
. C N O
- g 4 U g i _ - - _ _ _ _ _ - _ _ - _ _ _ _ - - _ _ . . 6 m V 6 3 m g [ = @ s 3 0 a e e 9 O C M N a C 30 % 6 4 % 6 M .
g .g ,. w e O O N E b
* ,
- W E
% W N t
i
,, g n o e o O N. N. N. N. N. N. N. N. N. N. N. N. N. . 4 G 9 M C ip g e* 30 0 @. O O O O O O O O O O O O O 9 *
- b i H D b g - y v v v v v v v v v v v v 0 6
- W@
C * @ % > * " E I 4 l i -O O - * '3 9 j g ,g y - = * @
*
- E O.e e 4 e E N E &w 1
g O @ 4 * *
=
3 g E e
% N.
N. t. N. N. ,. ,. N. N. N. N. N. N.
- O O p
e @ O O M C C. C 0 0 0 0 0 C C O O O O d W
- e $ v c - v v v v v 0 w- 4 ,
g I 9 % Q t O h C 6 3 6 e 3 4 e @ 9 C C J = c g
- O l e . , , , , , , , , -
, e e o o o E %
@ N. ,. . N. N. N. N.
ew " d 8 O
- O O. O O. O. O O. O. O. O, O O O O e > 4* F O e O % v v v v v v v v v v v v v v OO
>
- 9 C D C L y - 4 4 C E ,
9 e ( e oeo 6 t .C - l-
= = 4 Z '
i e @ 4 e 3 % E e N N N N N N N N N N N N N w t * [' E o O O O % N. - - - - - O 6 C O O e O O' O O v v v v O O O O O O' v O v O O v v O v
'i e a O = v v v v v C 4 6 = % 0 0 4 ;
& v = * * ;
- - - - _ -. - *, g w e - O C @ V C ,
O C. e e 6 Ch
- D C
. g -
C $ @ @ b e @ = N M 9 @ b e @ O 9 - D C s N N N N h M M M M M M M M 9 E 4 4h 0 6 *u l y > ** M
. e - N
= & 4 % = WW te a b B C 9 t U 3 3 O > > <
> O O %
.- - -- 2 2 2 ,
1 t I l' e l l .T l r
- - - - - - - c- - - g-,--y--w- y -yyww--meww,4,%y.g, ,gy -e._ _ , _ , _ ,_ _
p ,,,, p.. , _ , , _, , ., ,_ , ,, ,, ,, ., , _ _ _ . ,.
'A.
g C. ~ nr id
~N" E&
-. J2
*A W
l .
, _ . _ - = .
4, L 4 I l s
. . ..n,m .. .. : . . .n.,1 i 4
q m i 'l
- E iP's...# $*. .- -
p.- .,
- , , .,. A g -
3 a it I [ E4 [ {
- l. b"<;
m u -
.a v.:
i s ,. , d i,(J"
! I$ M,50ft 9. b !
; .. 'MNd
' d N0 p
; . iy I#$T@V 7.. I O! t,;N.it)h)*-4' .!_!
y I % $4 }3,"@. " u,ua .. p A hd s 7;.1 I i. fjg
..t .i\
6.e.w
'I-e
,p
<* 5.
%w r4 g i - >
t M'43 , ,1 k'g ;
, n !;. " - .
) % .] k L ~;I g1 t.A B a . .z . . , . w. gum e vg, I;
!d . !a . . /. - . _ _ . - , ,
.1
M
--- - - - -. - - p N
** N
@ e C e e N C
M M 4e .O. W E e N 4 4 6 g < N. O e e=
> D
@ O n
O M = O N y
. e. O. N.
O e O O O O
- e. N. N. e.
M C *
- O 2 O
N. O O O O N. N. N. g a 3 e C @
- O % v v v e v v v M &
O y NI _. - - - - _ _ _ _ _ _ _ - - _ -
-l e
M % W
- > s h
C O N M O
+8
- 0 m b @= 9 O e en N i N N. N. O. N. N. N. N. e. e N. N. O N. N. N. N. N.
8 - @ c = 0 0 N O O O O O M O O O - 0 O O O O C H D C- N O v v v v v v v v v v 0 C 5 h b O % ** y N 3
- 9
- - - - - - C g N .
- C _
& O E e
* *
- W W O % @ @ O e e e N O e N. N. e. N. N. N. N. . p . N. @. O N. N. N. N. N. e e- 9 T * % C O M C O O O
- N O O O a O O O O O O F D O M v v v v v v v v v v v v C N O
".)
- - - - - - - - - - - - - - - - - - - - - - .f V b g 9C g e 3 =
e, e e O M N a
- C W % b 4 i M
% N g *9* **
- 0 C N E h.= O N @
- W
- 9 g N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. 4 5 MQ g =3 4, a O O . . .
O O O O ** 9 f,
- W O @ O O O O O O O O = 0 C O M O 5 6 V N 4 - y v v v v v v v v v v v v v v v 0 b w a C - p % > - a =
b = Q o O = 4
,, O y -
e = a. es @ e e E G
- m e &
a @ g W e. D e. O b D w S 3 e 4 9 @ N. N. @.
- e. @. @. N. N. M. n. e <
= W L % . .
4 e O O M C O O O O O O. O O O O. C - O C O O O e #
- e y O = v v 0 w C C Z g % &= 0 0 m c -
- .N - - - - - .. . - g 3 en e' b 1 9 @ 9 C y e g .C. -
O C 0 E % 4 T. T. T. 4 T. T. N. T. T. 3 E
.e e e o O O @ T. T. T. T. N. T. T. T. O e 4 e C - 0 0 0 C O O O O O O O O O O O O O O m >
- O e O % v v v v v v v v v v v v v v v v v v @O C
> = T C a O y - e U
- . - - - - - - - . g n
- C 9@
- k
.C. Z
. - e 9 @ 9 4 3 e g e N N N N N N N N N N. N. 04 N N N N w e e s e O O so % + - - N. N. = - O s C 0 C e O O O O O O O O O O O O O O O O O* O* e e ea O - y v v v v v v v v v v v v v v v v v C J b - % e O 4
& y = 6 *
- ea e en g aC **
d - O 5 F li " g
. b
- 9e . 3 7
. R -
Q O
- N MW @ @ e= @ O s-A C E N M T @ @ > e e = * = e- = N g ( 4 4 O &
Q > .* U M n 8
- N 3
= & V = m e e > 0 4 WW 4 V 3 O j *
- A 4
> OO %
- . - - - -- -- 2 2 e 2
i-t i
- N
@ e C e e N C
O M M w e at E o = e w O . 4 4 4 "L 9
- e. &. >W o N N. N. N. N. N. N. % % % N. N. N. O O e V W O M =
0 0 0 0 C O 2 v v v v v v 2 2 O O v v O v 3 # W1 4 C @ @ 0 P-U
- . - 3 - - - -
5 m l* M %
- l. W N C - N E
- O P* M O a
*
- I > > @
N. N. N. N. N. N. O. e, o O N. N. N. e w N e N e G * -
@ @ a O O O O O O es O = 0 0 0 C
> O N @ v v v v v v v v v 0 h P*
- b U N 3 m
.u.s ui. n .mu. mm .mm. e - .s .mmu mus e.m. g N .
1 N C E O E e
* * $ $ 0 % N N N N N N O N N N N < 0
- O N O e + * -
- 4 G = 9 9 == % O O' O O O O* @ O M C O* O- O
> t @ M v v v v v v v v v v
-C 5 0
U O ! C 1 .. - - -
-V b g 9E - e e 3
# 0 M N a
j e e a
- C $ % b f I M g *9 *
- o O N E N
4 - 9 N
- e e e o o . N. N. N. N. N. N. N. N. N. N. 4 g M o 9
- W O @ O O C C O O O O eO J O
- w A >- > g - v v v v v v v v v v 0 6 WN
= @ % > = *=
4 O y o e = 9 O
- - - - - - - O E w 4 4 a w
, Q @ -D e- 9 3 e 9 4 M. N. N. N. N. @ N. N. N. - O b o p E % N. . A < 1 e @ O O M O O O O O O O O N O O O e e A = e v O = v v v v v oe C C 1 4 % @
- O 9
.b - - - p 3 =
c= t
- b @
.) 4 @ e C* *C 6 M ~ G & 9 C 0 E % 9 4 4 g 4 g g N N N s C i; .e
- O O O @ - -
- N. N. 9 4 i e O ~ O+ O O- O O O O O O O O O d D
- 4 0 0 % v v v v v v v v v v v v O O C 1 > - 4 C c o
.i w - eU
- e. .. - g
- e ef 8
g S
.O
= g -g I e
= = D A O @ n n 3 t5 E e N w 9 a
( g 4 o o e s . N. N. N. N. N. N. N. N. N.. N. N. O O O O O g g 1 0 0 @ O O O O O O O O w w 0 s O - v v v v v v v v v v v v C 4 b e- 6 0 4
& v .% 6 -
- 4 *
, - - - .. g w 4 @
- - 6 O
- D .C 1
e e. '
- N M 9 @ @ e @ O = N M & = C N N N N N N N N M MM M E 4 46 U > a -. m 6 9 mN
- 4 0 a d > C 4 W IW 4 U 3 C I *
- 4
> l .
O O % e e.a. ee .= 2 2 2
*E
$h 2n
- E8 mg, 3s
^
.a ..
]~ '
es h l ,
?EI _
s
?
tie n y
! l!
y 6 y = .. =. =. r:p g :
*: bVi 3r3 p! i.
; H{
<N
')-
W L 9,2 8 3 i:%
- x , ,
- $ 3.h u a y i; >
Ac,,,;,1 . m'" . m.. m.,$,g]. ' u.NM,.
- ~ ;1. . . .
~
"n
)1
?
g [4 % '$ a fl ,l !: j y} lj f ik. i I, d@n) s T; ' ' ":- l i j' U!Ip g 6 s@" f e
'i!A
, a1
- n.
>.i? -k r D @ O M e- O O O O O O M
- O O O O O O 3 4 C 9 * @ A v v v v v v v v v V V M CL I
- 6 O b O 1
. - - - - . l 6
9 .' M % W N N., C - E *- O N M O a e
- I k k @ @
e e N g N , N. N. N. N. N. N. . N. N. N. N. N. N. N.
*
- O @ =
O 0 O O O O
- O O O O O O O C
> D O N @ v v v v v v v v v v v v v 0 C P b .- b i 0 %
- i U N 3 9
- .=.=. - - . - C
~ -
1 1 4 s QE @ o % N N N N N N O N-I N 1 N N N N N g y *
- e N C e * - - - @
g e* 9 4 = % M O O O' O v v v v O O v v
* ('O-v v O
v O O v v O O v v O i
. H D @
a r g C p o t s V O l - - . - . . - .C . b i w . t . p U. 2 k 4 M N e M a C g .y . O O N = N e e=
, w
- N E e w e o o e. N. N. N. N. N. N. T. N. N. N. N. N. N. N. e t MQ J t = # C W O O O O O O O O O O O O O O *
- e t P. 9 .- y v v v v v v v v v v v v 0 h W 9 1 .g C - @ % > -- g -
t 0 0 D
,0 U =
.= ($ # ,
n
- O A
- - - - - - - - - E m 4 q
v
.O. CL C @ e. v 8 3 so g 4 4 4 -D o e $ s % M. M. M. F. , M. N. N. N. N. N. n 8 @ O O M C C C C O O O. O O O O O O O e d o *
. 9 y O
- v v v v v 3 g I T % 4-
, @ N C 1
- - - . g 3 4
, 6 8 @ e C d
l ee
=
s m
% < < < < , N
<. < < ,. N. N.
N N. f,
=
e e e O O o @ . . . . WD 4 e O
- O. O O. O O. O O O O. O C O O O Q 8 0 % v v v v v v v v v v v v v v O O
>
- 9 C D
= g
- .v .-. g 4 e 0 @ 4 4 j 3 t E e N N N N N N N N N N N N N w e s e O O e % - - - - N. o u O O* O O O- O O- O O O- O- O O o
o O O O = 3 s O = v v v v v v v v v v v v v v C k = % 0 0=
& V
- b 4
- 9
$ b C O D 0 4 e .C g
- D
. e
- T @ =
9
.e .
. U g [
- N M , @ @ s . O. . N. M. . .
9 O = N
- 6 e WW e e 4 >-
4 W Q 8-OO 6.- e - a. 2 3 1
~ , v,-w- r- e v- ~ wn-r-, nw,-,,e,--er,,-- - - - ,,--.--,An- , ---,m--- .-
40 C n yn
~
.A
. e MM M
-- a a w, .c ._r. . b -l ; .h l. . i CWJ -
/
. I!LL.iliI
_____ ____ .. r, i a 9 l i A p _ 7.s/. I;w-c-
- D E$ ll .
5 i p.! l b] s _ .l_E .. . _ ;s g 3 a di
- ; r- i f I rj .; ~9 R
E '
\. u.:
3 p [ n Q.- IN ' -( j y$".m;.; i! ,- r* d-- t nn
. . . t .. s .-___ g.m j.. .. .. , . . . . , A. , jw j -
.m. .t I "
s o ..d- * [x.3 U"]' ?m.v, .m.r. . . .h. 4.c
. Q3 .t1
}
a
- s. .
..A
-j n. .c .s b S a
-a ?, 4iD .
e' * ' , ), . q .,0$. ,o c I, yh
*N m - .
. n.. .j&w s,..-..
. _=. c I'@ ?;hTunuNf$. iif:qb Tb e
nf ll % ..... .. ...fi
.- .x
. E )
- g. ..-. ... .. e t j
- , = . . . , , , . .,..))
f . n_ . . . .= . c.. . - . . - - - -
I: 4
- - p P'
l w N 4 e C e e N C % O M w e $ E . M. G * @ $ 0 . 9
- e. k o N. N. e. N. N. N. N. N. N. N. N. $
> g 9+ O M N.- O O O O O O O O O O O e C @
- W % v v v v v v v v v v &
O N U e b e M %
$ p.
b C - t - 6 h M O . W - I N b @ e . N e g. . O. N. N. N. N. N. N. N. N. N. N. e= - e C= e O O O O O O O O O O C
> O N e y v v v v v v v v v 0 m > e b
% =
U 3 N. 9
* - . - - - - C g N .
- C ..
E O E e w $ @ O % g e =,= g N o g N. N. N. N. N. N. N. N. N. N. N.
- e* 9 9 .- % 9 O O O O O O O O O O O
> D @ M v v v v v v v v v v C U 0
U
', f.
4 U L 1 g 4 7 a g 9s e e O M N 1 I a
- C W b 4 M i
.'. gD* *
- o O %.
P M N N N N N N N N N N = N' i N L .= l
- e
- I O o . ~ - - - - -
- 4 i M 'O" O * $ O @ O~ O O O O O O O O O+ O
- 9
& > t P. g - v v v 0 6 W@
C
- c % >
- E
- b 0 C - 3 0 0 - - O e w . - O g -- - - - - - - - - - - - - - - - D E w 4 o C. c D o- w S 3 e N g M @ M 4 = 0 g a % M. M. . M. N. N. p 4 e O O M O O O O. O. O O O O O O e e
- 9 v 0 = v v v v v v v v o w C I g % CL -
> c l- .O - .e. - - - - - g 3 i e s i 4 @ 9 C l
~ e , ,
C
- U C e a % N , , , , ,* , , , i 9
d 9
- O O O t= e >
Q # C 0 C' O O O O O O O* O O' 0 e O % v v v v v v v v v v v O c
> - 9 4 y .C 4
- - . . - g 9 m
- 9 0 b 6 C 0
- O
= .- V O 9 e N N N N N N N N 4 4 w e L 3 t s e N.
N N s e O O o % < * *
- o s @
0 C O O O O- O O O O- O O* O~ O' w = 0 .4 0 - v v v v v v v v v v v C 3 h a % 9 6 Q
& v e -
6
- 4 es *g A .e =
C O V C 0 C 4 9
* - 4 4
- T S
- 4 =
- 1 O O
- N M 6 =
N g@ @ P. @ @ * .a *
- g it g g 6 9 = N D
* & = 4 e p- 4 W
> > g W e 4 U '
O O O 6.I -- 3 2 e
Reactor Building -4th Floor : 1 Elevation 659'-6" __ m m .v .3.. r..3 m ; - - ~ - - ~ ~ - - - =- 3
- 7 M 0""* 'h k --j-~
- 7 ',"*, . 1
+.,
( .g.wg:m-
.-- -- ;;:e,>.
.,.z, f.[/,.-
s I. ." k . .n >D**c '(o I e CONTROL 's 7 3.2 run exx. if T F [..:. s = f.,q
.y.
. Lm ,.,3' v,5 ci,f?,. 2:_ _ rry l f--
$F,Q5?Nld..ii.= m Nc.=.m m
+ N .
3 ?.5' I
.-. g. 3
- p g, + .-
Z. :<l l .b,e;. i Os Cecaps I ' u ? -- g #
~
Y. /j[ VENTILATION
~
MG Sees .
,z,,s .- EQUIPMENT AREA
*'.e.eny,.a't?.'
?.7f *.e. da'E. < " l1 ,
l . b-- I I -
" ' N, s-sm SLC c
P, 1
.< - ' 2C Puisun g . EM ,
Ava * ' '
/ .' T"T-T -
ul ,- I s.p.m. i- =g.~ I
*y % POu( g .
t .
.] p
[ U "' il. '
;N jay 1
a.3+ n E,e c :3.m.:c, ;: .:- = :..:.. n. . . . . .
- .... - ,... =. .
! pg.% t 6m A. l.;>-c i r w Ae % 1 (FIGURE 3.2-3) Supplement 9 (Page 17 of 32) Page 3.2-28 I
/
w N e e C e O N C % O M M
.=
w - a. E = O e O O e .* g b. N O N N. N. N. N. N. N. N. N. N. N. N. N. N. O O M - C O O O O O O O O O O O a. C 4 eD * % v v v v v v v v v v v v v @ CL C @ 0 6 Y m w M % W h
= h E 5 M O a e e e 6 m e= e N o. N. N. N. N. N. N. N. N. N. N. N. N. O. N.
t
- N. @ @ = 0 0 C O O O O O O O O O M C
> 9 O N @ v v v v v v v v v v v v v C 5 6 = b o %
3 U N
= 9
.- - - - - - - - -- - - = C N +
C k E 0 E S N N
*
- W @ O % N N N N N N N N N N
*
- N O 9 ~
- N. - N. - *
- S 9 - T = % C O O O O O O O O O O O= O" O-
* > @ M v v v v v v v v v v v v v v b p e __ _ _ - - - _ _ .- - _ - _ _ _ _ _ . . .
s
. V. -
o .r w
.8 s 4 nN e M Ag.. *
- C o
N E n= N N =@
% N
- t.
l I
> 8
- e O O a N. N. N. N. N. N. N. N. N. - N. N. N.
4 m o i e* W Q W C C. O O O O O O O O O O O O ,
> b g - y v v v v v v v v v v v v v 0 b WG i g a @ % >
- Cr =
=
g O O. = c U * - e Ch
- - - - - - - - - - - a g S g
- CL
@ D u. w
@
- O
= W w % N. N. N. M. M. M. N. M. N. k. N. N. N. M. W 4 @ O O M O O O O O O O O O O O O O O e e v O y v v v v v w T %
, @ C
, .- - - - .b - - - - . - - g 3 6
S @ 9 g - . , , < <. , c, .C d c .e C O o s %
,. ,. . M. . N. N. N. N. N. . e
- O - 0 O O O O. O O O O O O O O O d >
e O % v v v v v v v v v v v v v v @o
>
- 4 C D y = 4
- .- .-. . - .- - .- - -- g 4 e b
= ** V I
@ 4 4 t
s e N N N N N N N N N N N. N N w e s O O o % N. * - o s t e @ C- Cv O- O O O- O O O O- O O O O w es v v v v v v v v v v v v v b o % $ v 6 .
& =
4
- j
.- - . - .- . . .- - . . . . .- g I W =
. . s
- O.e .
@ O
- N M M e m C
- N M 4 @ @ h 9 @ * * * = = E 4
> . =
U w N 8 9
* & C =
* > 0 4 WW 9 U S O > >
t---- _--------_---- 92
a J n i6 2" m2m ' g&
.. J2
. y m ...,, ...
h
! I, 6
- - #l 8 t f 5 Imd
- Fy !,4i,l l
; Li ia i i P e
# oja X
'L -
h%,. , . .' l
- m. m . ... , ,a 9
N.$$g$$\,s..m$b Y
!?lli T
I A
$g ~
i o e
.g c.m m . . . m . . . . . .. .. .
oe am= -_ z s.>ea. au { n' l , . I r--Y' ~ e aa
] e *" !k } T qu fi D
81:' O j i I! .!lE [y r y 3 d._ l* ? e ~ r j
.~-~~- y ti
.l '% Y
- i. r4 m . :g g
$ if,
% L
# ?,
.l S'N :
.1 li e
dh3 IJj %f lx l: hh v 4 i.
- z. . . . - , .
I i 6 i
. ,- , _ . - - . . . . - . . .__ . , - - - - _ _ - ~ . - , . , . . . . _ _ _ _ . , , - . _ _ , _ -
_,#d,_ _ _ _ _ . r e
- M e e C
- f e N C a M
O M g w - d E g @ O e. @ $ g ,e &. > o N N. N. N. N. N. N. . N. N. N. N. N. N. N. N. g eg 9 O M = 0 C O O O O
- O O O O O O O O e C @ e @ % v v v v v v v v v v v v v dL 0 b U .
e M % ' W k C - k E = , O N M O = w
- I N N @
g a N I N 1,0 N. "e. N. N. N. N. . N. N. N. N. N. N. N. N. ' 9 == 9 e a O O O O O O O O O O O O O O O C
>= 0- N @ v v v v v v v v v v v v v 0 6 M .* 6
% es U N 3
- 9
- - - - c 4
N + " Wj c > . o a e N N N N N N N r w - @ o % N N N N N M N , e
- N O e - - - - N. - - < =
0
- 9 * % O O- O O O O O O O O O O O O- O O t H @ M v v v v v v v v v v v v v v >
e 0 ' 1 O g V s V b a l g * * @ 4 3
,9 e WO M N b 4 I M m
- C W %
% N
-[ g O O N N N N N N N N N N N N N N N N N = W e 4. 4 *3 w e 4 s= MQ S e
- e C O = - - =- * - -
4 f
- e S
- W O @ C O O C'O O' O O O O O O* O O O>
j 9 > V > 9 - y v v v v v v v v v v v v v v 0 b WC C C = @ % >
- a N
.+ e g p J y o. e =
- a. o w e a f
- - - - - - - . - -. - t 3 m g I > a th D s w
- O @ '
j 3 e N = 0
=
- g a % N. N. . N. N. N. N. N. N. N. N. N. N. N. N. g C 4 @ O o M C O O O O O O O O O O O O O O d
- 9 v O = v v v v v v v v v v v v v v v o * .
t .O I 9 % 4e 2 j
* @ > g 3
+
g _ - - - - - - - - { i b 4 y @ 9 C= w .6 w - e t. 3 r C 9 & % N g N. N. N. e o o o0 4 N. N. N. N. N. N. N. N. N. N. N. , 1' . O O O O O WD e - O O O O O O C O O O t + 0 0 % v v v v v v v v v v v v v v v @O L
> ~ 9 C D t
- e 4
s - .,,. .v - - , , , - y I 4 e i e o b Ce p m i e @ $ . . 3 g s e N N N N N N N N N N N N N N N w e i' b" s o O O O % - - - -
- o b e O p C' C O' O- O' O O O O* O* O- O O O* O' e (
ew O
- v v v v v v v v v v v v v v v C g 6 e- % e o i
& v
- 6 -
8 * > e.> s.. . a.. i i- s . .w m.. e . $ f ] d
- 4
@V [
C 'l O C 4
* - b j
- g .
- g '
O
- N M -4 *@ -@- h t = -
1
- N M 9 @ e e . * - - a . .
$ '9 ** t Q b
- N T 4 $
a es & w g WW e > >= >= ] li 9 U Q _a____ -________-_____ ?? I i f l t t L
N ed M P c e k ! b N l ". Ji 4' 3" ae Z 3" s? u e,,' -~ g V .:..' A A ..' "at yj ' C 5 7 ~- - 3 P,
?
I
~U
- , .A - > 1, c
4 .k8 g"q i ) di
.g i
'w
'%N.',I'f *_I,I .
'% 3', # g f. l< J W
,(l w ,4,,
8 2 w N,. . 7
=
ij f .a .
. 1 L i
B *
$'5] i ,
[f.s, w- . u . . . .. lglbpa J c e g 2 9 O- ~5 -
/ J
]; - - ,
en 1 i;,,,t ;: " " a-t .> ->
, h; ,1 J :=l d
. ss . ..
. . .~
g
;' e y
$, m y, w i
. Q.]g ~
[~ w ' # ~
k .VM yM y 2< ;.. . i ,5 ; :. e
( .q' I-L . y - ('.,m 3 %
. e.; __
b
. C=
3 , wi l *I *1 1 1 ,Q .' -
, s g
I, ?5 '5_ o t'.$YI M' ~i J ,P l Y s/<
* ,e_..
r
. ;. - . m 4,
- e
% ,- 3 q t=F,. * ,1
, -< w
'c 3g _
a 9.ggQ: % & . >=
y, _. ._ _ .._ l l l i d J J s
-.- -- - - - . - - - O M
- M e e C 8 e
% N.
C O M e e 8 k M.= a w $4 W O . N N N O N N e N N N N* O N N* N N N 9 N O N * * * * -
- 9 * * <
> G O M = 0* C O O N O* O N- O O O O e O* 6O O= O 4
$ = @ % v v = v v v v v v v v v v v v 4 N i o . !
e M %
$ N c = m a =
6 h M C =
*
- e b h 8 @ @
e . N e N = N. N. O. N. . N. N. O. N. N. N. N. . N. N. N. N. N. i e*
- Ob it: - O O N O O O O N O O O O M O O O O O C
> O N @ v v v v v v v v v v v v v v 0
> N
- b U N 3
= 4 l 1
- - . - . C N .
g h !
. C g o 4 e :s N N N
* * @ G % N O N M N N @ N N N N N N 0
- N O e - N. = * * - - - 4 y
g 8 - 9 *- % C C N* O O O- O- O O O O O
- O O- O* O O* O v v v v v v v v v v v v *
> y @ M v v ,
s a O 8 ; L g
- .i - - . - - . .C . M r a t b 4 m g j e @ @3
, , e e oW M N i
m y C W % l b d 4 I M
- g. O O ' r. l % 1 N
, W
- N E = N N N N N N N N N N N N N N N' N N N = e @ *
- s * = e O O - * - * * - * * - *
@G MO >
g 9 - W O @' C O O= O OO O' O' O O* C O O O O O O
- 1 #
> t > g = v v v v v v v v v v O. v v v v v v v c 6 W '
WN -
= % > e C N [
g a C @ = *
. O O J ,g y =
- 8 mO 9
l g - . - - - - - - - .. - - - e E [w u. 4 , g = a p 3 C. @ a e. O w t 3 e N 6 - OM i w g N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. e ; W % - O O e W M O O O O O O O O O O O O 4 @ O O M 2 O O O v v 0 ** %
. 9 v c = v v v v v v v v v v v v v v v v 1 9 % O l @ CW
- .N - - - - g 3 4 b
@ @ e C U l
C .
=
s % e N N N N N N N N N N N N N s. N N N N t, e O 3 a t
- O O CO @
- O O' O O O
- =
O O O O
*i O* O O' O O O O W >
a e = 0 v v v v v v v v v v v v @ O E O 4 O % v v v v v j
> = T C a O
- e o t
.v - - - - t a ;
4 e ! e@ = 4 C 4 ?
* [
=
- ow
@4 9 @ 9 +
N N N N N N .N
- N* N* N N N N N N = $ 7 3 0 t e N N N f s 4 O O O % + a * * *
- O b @
o o @ 3 O*- O O' O O O* O* O* O O O O* O O- O- O- O
- I e s O = v vi y v v v v v v v v v v v v v v v C C b = % 1 0 0
- y 6 y = b =
@
- s
. -. - . . - . < g O
e t b
= e = 9 ,
O = N M T @ @ 6 - E i
- s. N M g W @ 6 e@ = se = s= == = = h.= e. a <
h y > .. .. g e t = N > 6 a e & C
- e > p g W w 3 9 U u O I > > H $
> l
- f. -.-- - e
- - - - - - - - - . - - - - - . - - 3 4
- M
@ m C 6 CD $ N
% E
- C O M $ M w * # la - -
Ww @ @ O . N N. N N. N N. O. T N @ N. N. N N 9 N ae e
- h C N . . - -
a O H V t@. C M a O O O O O O - O @ C C - O C C C* * - 3 4 C @ ** @ % v v to a O N O m e M %
# P=
C - 6 s - O h M O a w ** e S b @ N O O e . N e N N. N. N. . N. M. N. O. N. N. N. N. N. N. N. N. 9 = .
@ @
- O O C O C t,,, e C @ C C
- O C C O O O C H Q O N @ v v v v v v v v v v v v v v 0 C N b = b O %
- O N 3 9
- - - .-. .= - - - - -. .-. - - - C b ~ .
g
. C N g O E e
*- 30 @ O % M
, W ** @ N O CD N. N. N. N. N. N. - N. O. N. N. N. N. N. N. N. N. N. 9 y e - 9 9 - % C O O C C O O O N O O O O C C C C O O g D- C @M v v v v v v v v v v v v v v v C et id. O V O g
C V b a 4 3 a g F @ 9 0 f4 N . . tD a C p b 4 i M g .g O O %. P N N
. * - N g = N N N N N
- M e t.
y 3 e e e o o . N. N. N. . . . N. N. N. . N. N. . N. N. N. N. N. e E n O g g e == # o C C O C O O O C C C C C O O C C O C
- 6 3 S C 6 9 J.
e y v v v v v v v v v v v v v v v v v 0 4. WM e C * @ % >
- 2 N
=
k. g .k g U O O
- 3e
= 0 g - - - - - - - - - - - -.. - 9 8 g g
- 1
,3 a @
N e. c e. w 3 CD N N N N - O
- a. p w % N. N. . es. N. N. . N. . N. N. N. N. N. . N. g 9 4 O O M C C C C O O O O OO C C C. C f) C C. O W d 9 v C
- v v v v v v s y v v v v v v v v v v o e g I 4 % Qe
.,. - . - - - - - - - - - - - - - .-r -
o 5 V - e w g ae i O g g g s 3 g g g g g g g g g g g g g p, g g ,
=* q e o o O @ . . . . . . . . . . . . . . n. .
e a e C - C 0 C O C C C C O C C C O C C C. C' O e D O 4 C % v v v v w v v v v v v v v v v v v v 0 0
> ** 9 C D v - g
- - - - - - - - - - - ,. . . y - - - - g e e 9@
- s. C 0 @
to P- N ,e 9 e 3 V g e N. N. N. N. N. N. N. N. N. N. ee. .' N. N. N. N. g g c c e s . . C C C C O O C O 6 O C @ C C C O O O O O O C C
- 9 .J C *
- v v v v v v v '
Y v / v v v v v v . C e m % $ 0 A v - 6 - G d'
- - - - - - - - -. - - g e
C O g O C 4
- - 6
- D
. g .
V ^
@ C - N e 9 @@ k oc @ O ~ N M 4 # @ e =
C a. N N N N N *s N N N N M MM M n M M g at 3 a L' t a= N
** 4 4 T
- WW
* > J C 9 9 U 3 3- O > >
> i O C
.- .=
- . - g g
5 r t L wM W e C 8 e 9 N C % 5
- O M 6 M
==
w - d E e-G 9 0 w @ O O < N N N N N e. O 9 = > C N
- 9* ~ 6@
> D C M == C C- C- C* C C C C- 3 4 C = @ % v v v v v v = M &
O N U <
- - == - .-.
e M %
$ > 5
=
C - t ,
+
O > M 0 =
** - I k N @ !
s e N s N a= N. N. N. N. N. N. N. O. e* -
@ = C C C C C C C 9 C t
> C e.
f O v v v v v v v O L I
> > e en U N 3 e e
- . - C ,
; N .
y k
. C g 0 E e w
- D % N t. ce N N N w 0
- f.
r O e - - N. -
- N. -
g #
- 9 * % C C C= O C C C C 6 > @ M v v v v v v v v n
e U @ C p g w . f .. 1.
- i. M N 8 M q C $ %
h O C - N. w } .e w - N . - c. c. c.. e. e. g M o
.g g d w a c o . N. . h. N. . .
ip g G * # C @ C C C. C C C C C ** W 9
> T A 9 - V V v v v v v v 0 b i
% H = N
.j j a C s- @ = I { O O == 9 ' j @ g U ==
*
- m@
t
-- - - .- - - - e g S e ; ! g 4 j 3 o @ "c -
i
. p 3 co N -o g
L + = p a % . N. N. N. N. N. N. N. 4 @ C 0 M C C C C C C C C d e e 9 v C e y y v v v v v v o e i-g Z 9 % @ e
@ C I .M. - g 3 g g e, .-. - - .-
b 5
. g @ t C i
l
= , e r. =g l
** e . . - c. c. c. c. N. c. c. c.
e d co o @
==
C C C. C. C C C. C d > ! 3
- O v @o O e o
- v v v v v v v
> ~ 9 .c 4 w
- 4 i.
i
.= .- .- - - .-- .-. g j e e I
, 9@ s b C t J = as so Q b
@ 4, 4 t
]ee In, C C O s e
- c. c. c. c. c. c. c. c.
w 0 % C- C' C- C- C 0 C @ C- C ' l es 6 c v v C. v v v v v e e C
- b e
& y as 4
- 4
' .se - - - g. .W. eI - - M - g g d = =
i 1
.O - . .
i
- N [
i - e ,
- c. 0 =
i 9 h e S C
- M 9 j 3 M M M 4 9 9 9 9 E 9 6 a =
U > *= N 1 9
- O 4 C'* WW f a w B o 4 > >
4 W X f C \
- . -c - - ------ - 21 i i
i ; i ! t-I 1 f i t 1 i ) l 4 i i I i I
-y.d. - 7, , ,... - - _ _ . . _ . , . , , . _ , . . _ , _ - - , s,-_._.~m,,_- r~--,rryn.-.__-m.., m., , , ,y., ---- - , - ,._ ,
C l 7 !"g CL
]
,_.) e .t 1 h 3 kj lI l
1 w wn(c a WI t I e raje r.
, , . . . . . . ., 7;:
- Q C.; 3] j, e 2
*piw- s . s --
. t.
3 [
, pp(q-iAe,i d3jW,A
- c. . ,
- q ,, .
A
}- .'! -e -
- <i;.
-M 3-'.gC,d
_ ._ . }? [5 ![ WR 5' '? 2 j '_ .1f j. 4F -
'. <% A. i 5
b1 y l- {frG4m 1 ;*
@ ~
r L. W4lj iy:..y. .c2 s i 5... r tt 5g * ""('r" rf: a# 2 0ti*.3 rts 63 _*NF 5 5J- , 3 7, tj'g($. p.3 3-4?! f;:]q%
"l l.
3.? w '
- 5. f 5 'w i,l h s 3.
h,1 aa$ T+&, kC hh m E *
.g g ,. ;,=,. -: , y . T
- a; g
5e l '. lu ?.s .: s .n...u. p. ) x.} i 4 4
-- . - - , - - - . .-.-e , ___
. - -, . . , --.. . - - , - . - - . . n.- - , -..
l l l l I 1 i l
- _ . - -- _. _ __-_ e N
& n
@ @ C I CG 9 Ne C % t e
w e o
# E e -M e e @ @ O . N N N Q 9 y e n o N . N. . . N. N. N. N. N. N. N. N. N. N. N. N. o O
> V W. O m - O O O O O O O O O O O C C cc 0 0 0 0 3 e C @ e @ % v v v v v v v v v v v v v v v @ v v M 4 0 h O
N M %
# 6 C - N s
- O N M O a e
- 8 h N @ N N N a e N t N e N. N. . N. N. N. N. N. N. . N. N. g N. o N. N.
9 = @ @
- O O O O O O O O O O O O O N O m O O C
> V O- N @ v v v v v v v v v v v v v v v v 0 C N N
- b 0 %
- U N 3
- 9
-- - - - - - =.- - - - - - - - - - - - - - - - - C L
0 N + 0
= h C
% O s @
N w * # @ O % N N N N N N N N N N N N N
*
- E N O @ - - - - - - - - - - - N. N. - - N. %
e - 9 w - % O O O O O O O O O O O O O O O O O O O
> V @ O v v v v v v v v v v v v v v v = y v W e C
j 0 V O 1 - - .- - - - - - - - - . - - .C . V b g 3 ^ g @ mN [ @ 9 0 b f 4 M g C # % g C " w e 0 C N w k
= N N W N. e.
y = n e a o o . N. N. N. N. N. . N. es. N. . N. N. N. N. N.. N. N. N. 4 g m a e* # 0 @ O O O O O O O O O O O O O O O O O C
- w y
y j > V 9
- v v v v v v v v v v v v v v v v v v 0 6 W@
g C N.
. @ % >=
- g N s 9 C C
- D g U e
. .- L3 $p
=C C - - - - - - E a 9 4'
=e {L 3 C ( C v. w
> 3 @
- o
= p a w N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. e e @ C 0 n O O O O O O O O O O O C O O O O O O * *
- 9 v O = y v v v v v v v v v v v v v v v g y o C Z T % 4 .e*
O @ h C
= - - - - - - - - - - - . - - - - - g 3
- b 4 # e C V = e f. -
0 C e a % a
.- Q e* o o o @ N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. N. s
. O O O O O O O O O O O O O O O O O O O e >
0 6 0 '. v v v v v v v v v v v v v v v v v v tp 0 C D
> e 4 - e v
- - - - - - - g e d 9 0 h
.C M
m e @ e 4 3 V a e N N N N N N N N N N N N N N N N N N = e
- u. 4 O C O % - - - - - - - - - - - -
0 % o O @ O O* O O O O O O O O O O- O O O O O' O e v v v v v v v v v v v v v v " v C ea c-v v 9 0 6 b e ( y 1 *
- - - . - ., - - - . - - - - - g d -
s
** -C
. e o G. Q .= N m 9 @ @ h C 4
- N.m g #@ h @ p - s. - - - e = =
- g 4 a s.
4 y >
- N I
- CL $ V .$ ww a F B e ,a
& V 3 3 O 0 0 l > - -- -- g 2
i
'y
= ,
1 [
- -- m e a ;
l w M j $ e II i e . i N e
C % i M
> O M e O O O O O O O O O e O O O O 4 :
- e % v v v v v v v fi w = v M v v v &
N b U 1 k ' .6 M % ! 31 h C *- N u t O 9 M D a. t w e t > > N l e w g N e. N. N. N. N. N. N. N. N. N. O. t .- O O
- e= N. A e
- O O O O O O O O O @ C N N O C r
> 0 N @ v v v v v v v v v v v 0 I N N = 6
- g e Q 9 3 g
. - - - 2 y N . Ib O E E w at e 0 % e e, a N O e N. N. N. N. N. N. N. O. N. N. N. N. N. N. l . O = 4 9 - % O O O O O O O @ O O O O O O O , W e e M v v v v v v v v v v v v v G t l i Ul' V E. 6 ! 3 a ' } g = @ e 3 , , C e 0 0 M N * % b 4 e M 4 j %* * - h N & h
- N N N *
% N y ." e e n O O . N. N. N. N. N. N. - N. N. N. + N. N. 4 MO 2 ] t = yt C @ O O Os O O O O
- O O O O O O
- u > b 9 = v v v v v v v v v v v v v 0 k WN
- e % F
- N t
= i u O o .= 9 O ( g -- - - - - - - - - - - - - - - - - - - E E 4 _D 4 3 a e a v. w k 3 e
- O i
( ge s % N. N. N. N. N. N. N. N. N. N. N.. N. N. N. e ' e c 0 0 M C O O O O O O O O O O O O O e e
- t v O = v v v v v v v v v v v v v v 0 * ,
{ 1 9 % h Q* C w @ I , . - - - - - . . - - g 3 e b e e a C y . e t e e C C &
- N N N N N ** N N 3 e
- O O O @ N. N. N. - N. N. N. e er O
- O O O O O' O O' O O. O O O O O 8 >
- t O % v v v v v v v v v v v v v v nO 3
- 9 .C e D >
v .e - - - .. - g e e i e@ i y b C , O @ eg 1 1 E e N N N N N N N. N. N. N. N.- - N w e s3 I> O O o % - - N. N. - o 6 ! c Q p O O- O O O- O O O- O O O O c o e i es v v v v v v v v v v v v v v a s o. v .5 6 - 3 t es I - - - . - - - . . -. e ; e* 'I C @D e * .O .C 6
- D g . ,
o O - N M < e e e O - N . . i lC N MM a < .-f y > l - N N N N N N N N M I e e = N * & 9 Tc e. ! e > ] 4 W Ink *=
- w q 4 U a 3 O
- O O
- - -- - ------ --- - 2 2 k
i F h t a f I + a 1~ m <l L~ _ _ _ gq n 2" 3a ,3_J ! l i.se , j . s w < ,r3 < . ;n : i *'}}! w , - 1s a ., l I f- ~ 4 w1 w a a 6 ) g e h I id . I NI C f 2 I; ! es I k' N_ f'!I*l w, . ,1n 2E ., .. a\ . .# ha . * * 'd,e 'l $" g'( . ' su:t--h coupo ,t 3$ ;2 ~~ 6 u wo Y7k l I
- Uj.h? iN_' ' ; .M' '
[, * ",7 " ' . 's M j t ! 40 . . . k . .*! , ., h %) - 5 t (a1 4 m - R. 1 iTS $
- i n %(! $c= : El ,
l U h?/h < . . .{. m m. . ..-,4, , _ _ _ . _ _ - _ _ _ _ _ _ . - - - _ - - c---m - - - - - - - - ' - - - - ' - - ' * - ' - - ^ " ~ - .- - -- -- vi t C5 C I @ 2 N e
- C %
M O M w e # E e N N N N N N h N '.'. C6 e a w @ co . e. N h. N N 4 *i Pe * -
- G@
9
- k O N * - - * -
>= c O fa
- O' O O O O O = 0 0 0 0' 0 0* C O* O O- O 3 4
@ == @ % y v @ v v v v v v v v v v v v et & U h e M N sf h C *- h s e o > M e a N N N N N h N N N w
- e >= h @ N N N N O h h h e e h e N e .'
- O - - - - - - * -
t * - e e a O Cr h O O *= 0- o- o- O- O O O+ O O O+ O O C > t h e v v v v v v v v v v v v v v v v 0 6. (O O. P k w U N 3 0 - - .=.=. - .-- - - C h 4 . ~ & c g e. N ed *1 / @ O % N N O N N h h N N @. N. N. N. N1 e *\ N O E ~ + - - ~ N. - g O E O O D C- C O O O O- O O O O O O-g ee g - % C v v v v v v v v v v v v > M v v v > w @ g - - - - - - - - - - - . - - .C . C L* m @
- q [
w . e t e 9E b 4 M h 4 M g g C pl % % N g O O h . m - N E == N r. N N N N N N h h N N N N N * * *. * *
- O t1 ~ - N. - - - - N. N. - - -
- e (* O g g s 4 == W D @' O O
~ O' O O- O O O O- O O O O O O O O O * > t h g
- v v v v v v s v v v v v v v v v v v e m W@
3 C * @ % >
- a N 6 a = 0 gs .[
J C O- a- Q t V * * . O g .- - .- - .- E w 4 g =3 g a e a v. .&. , E N N e. N N e. N N N N N N N N h h N N
- 0 3 d
- af & % * - + - + - . - -
e c C 0 M O O+ O O+ O+ O O+ O O- O- O+ O O O O O O- O- e W v O
- v v v v v v v v v v v v v v v v v v
[ J . g % & .=C @ g 3 ., - - .h . - - - . . - b 4 @ 4 . .C y . N e. e. e. e. e. e. e. e. e. c, d e e s % e. e. N. N. n. N. N. ev. g O o C . . . . . 6 e O . . . . . 3 e O
- O O O O O O O O O O. O O O O O O O O.
p o O t O% v v v v v v v v v v v v v v v v v v >
- g C
.C g - .vm.. .m. .= e-, e e e m .-. am e-. .m. g e d 9@ h C e .* g a e e 4 3 C 4 s N N N h N e. N e. e. N N e. N e. h N N w e. N. - - - ~ o s s e O O O % - - - - - . - - - 0 O e O ~O ~v OO v O O C v v v C v O O O v v v O v v O v O v O O v v O v w es 0 == v v e o C 6 & v
- b =>
G * == .- .. - - . - - - - ... . . .- g * =- C S V O e * .C g en g I e t
- C
- N m e 6. e e .=
C - N M g c @ h e@ - * - = - c.- @e - - a .s 0 Y p. om Pg 4 g * & G V
- WW
* >- E C 4 *= >- $ U 3 3 C l O O - .- -- --1 ., i - -- -- g g i h L see t I w 4 e e C
- N et
- C %
O s M M = = W @ 0 =- N N N N C N N N @ c 4 M N ~ e e @ > C < N. - + - - - - - C 9 e* O M N O O O O O O O O @ O- O O O O O O- O OO k t @- a O e v O v v - v C v v v v N 3 e C @ % * * '
- O ek 0 b V
.w - - - - k C M % C, W = k E F.* e O N M 0 a * - e > k @ h h O N N N N N N e
- N 1 N == = 0 N. N. O+ C N. O~ N. - N. O G == @ @
- O e O O @ C - O cO O O= C+ O O $ 0- O- C k O+ N in v v v v v v v v v v v v v Q k k
- b V N !
I e - - - - - - - - - - - - - - - - - -] - - - - - N + 8 e - g 0 E e *
- W VI o % h O N N h N N N N h N N N N O N N q *
- a N O e - - - N - - - - - - - -
G 6 4
- 9 9 - % C N C O O O' = 0 C O O- O- O O C' M C C O v v v v b
}0 @ f9 v v v v v v v v g - - - . . m D L e n Q ^ q [ @ 6 O M N e . a3 # % b K i M g g C % g . O O k - W N. e-m
- N g a= N y 3 e e s o a . N. h. N. N. N. N. . N. h. N. N. N. N. N. N. N. N. h. e q n o g g 4
- W O @ O O O C C O O O O O O O O C O O O O
- e y b t k g == v v v v v v v v v v v v v v v v v v 0 h WO g a C *- @ % b
- E M
= D f ,[g u 0 0 e = Q 9 w O g - - - - - - - .-. - 0 E g e = 4 ,3 C e C e. w 3 m h N N N N N N - O e p a % N. N. N. . N. N. h. N. N. . N. . N. . N. . m e c O O M C O O C C O O O C O C O C. C C C O O * * ) & v O e v v v v v v v v v v v v v v v v v v 0
- g I g % a*
. c h - - g 3 C i 6 ' E c e C f C 4 m t N. h. N. N N. N. N. N
- h. N. N.
N h N. h. N. N. N. E
- 3 8
- e e q O o @ . . .
e > G 8 C - C O O C O O C O O O O O O O C C O O C e O % v v v v v v v v v v v v v v v v v v 3 0 > a 9 C 0 - e - .v - - - - - - t e a 9@ 6 C e . t t e e e 3 t s e N h N N N N N N N N N N h N N h N N e e E 4 O O O A - - * - - - - - - - * * * - O b l o O @ O O O* O C O O O* C O O+ O C C O O O O- w O .J C ==
- v *
- v v v v v v v v v v v v v v C k .- % 0 0 4 y = b e 4 *
- - - - - - - - - - - - - . .. - e C @ t 0 e .C 6
- T e e.
T C @ O.
- N m 9 # @ k # @ O - N m 9 # C e a
= N N N N N N e N N N M P M P e mM g 4 g
- y > .
e e - N
- e. & 4 t
- WW d > B o e *
- t . V 3 3 O OO
-- --- - ----- - - 2 2 m _ _ - _ _ _ _ _____ _ _ _ . O , l w T g e C e e i N 1,,
- C %
O r 'l M w - # E = m e @ @ O a N N N N N N =CL 9 ft - N O fv N. * *
- Q@
> c @- O m = 0 0 0* = = 0 0 3 4 C @ * @ % v v v v v to 4 o N U k E M % W A b C = s = 0 h M O a e e e b m @ N O N 8 e N I N = N. N. N. - - - N. ee * @ @ = 0 C O O @ C O L > O' N @ v v v v v v 0 = f. & b b w O % U N e .. a - - - . N 6 8 C ~ g 0 E E *
- W @ O % N N N N O N N.
e * @ N O
- 4 q 8
- T 9 *
% C O O O
- O O O
6 M v v v v v v > D @ a g C > 0 " ? , - - - - _ _ - _ _ _ e a V L a g { = @ 4 3 a 9 0 M N . . b f e M a C W % % N g .y. *
- O N
O k = N N N N N = a m- . O e e s O O = - N- N. - + E {e M+ p 3 ** g g 9 = # O @ O- O O O- O O O > 7 P- 9 - v v v v v v v O b uJ == p >-
- a M 6 g C ** @ %
S 0 0 = 0 g .C U = = 0 9 e e = O g - - E 6 4 g = =9 a G @ D e. w $ 3 2 N N ** N
- O w s % N. N. N. g W - . .
e e e @ O O M O O O O O O. O e e s o a v v v v v v v oe g 1 %' % G= @ h C ., -. - - - - - - - - - - - g 3 b 4 @ 9 - E t .C l t e E % N N "o N N N a *
- O O D @ < - . ~ ~ N. 8 e C
- C O O* O O O O d 3 D O % i w v v v v v OO
> = 4 C D i - e - - - - - . - - - g 4 e 9 O b C = . . t e @ 4 e $ D a e N N N N N N w g h e C O O % -
- N* - O w o C O O O O O- O O* O* w 4 s O - v v v v v v v C 6 e- % 0 e
& v = b e G * .- .- - . - - - - - - . e 0 = C O T1 0 e .C.. 6
- V g .
=Y h e @ O = N M e a C M M M 9 9 4 4 45 4 0 *= Y o-e 4 = N
- 4 6 t -
0 > R C R se G t U $ 3 O = ** > O -- -- 2 M i7 i _1- ~ .... 1.( Z 3 *i *~ ~ ; *n =% - (i L h r f.,I. *} < J 4 O li .: J c . -- , . 3 ! . 8./ u..(. . s. ' # j!, l :nnQQ6s.J l ;, AY 3!fv.Nb i i NI aE . W_w fm't l t~ I' , t i 0 n y A ' l iQ i' k if I 3& I Yb ! ei , b, c. , bh13o,.1.,;l[jk). I n _ . - l ~ll h. 5 -. ~, ~ , - 5 F = ' n . 37 as - : ! 8- .us3. t "< , 51 e ,, . F. l\1
- r j *j tW h-[lIa.e.M..;=r'JlRl m .. u J i i Is.u u#:
d Supplement 9 Paga 3 3-1 l 3 3 Fuel Loading 331 Purpose The purpose of this test was to load fuel safely and efficiently to the full core size (764 assemblies). 332 criteria Level,1 The partially loaded core must be suberitical by at least 0 38 percent delta k/k with the analytically determined stroi. gest rod fully withdrawn. There must be a neutron signal count-to-noise count ratio of at least 2:1 on the required operable SRMs or fuel loading chambers (FLC). The min!aua count rate, as defined by the Technical Specifications, must be met on the required operable SRMs or fuel loading chambers. Level 2 None 333 nasults Prior to fuel loading, all fuel assemblies were inspected and then stored in the fual pool in such a way that no rotation of fuel assemblies would be required during their transfer to the reactor vessel and also that no assembly would pass over any other assembly in tha fuel pool during fuel loading. The only exception to this was bundle LJK 954 which was oriented SW instead of SE in the fuel pool, but was verified to be properly oriented in the core. Before the start of fuel load, all control rods were fully inserted, all blade guides were positioned as shown on Figure 3 3 1. coven sb-se neutron sources were installed at locations shown on Figure 3 3-1. All applicable initial conditions were verified prior to the start of fuel loading. Four times during the fuel loading process, fuel loading was suspended for greater than eight hours, and all applicable initial conditions were reverified bercee fuel loading was resumed. Suppler:nt 9 P:33 3 3-2 , The Botton head drain temperature indication was used to obtain the Reactor Coolant Temperature at least once every eight hours (1 15 minutes) during the fuel loading process. Detailed fuel loading sheets, approved by the l Reactor Engineer, provided the instructions on each j individual fuel assembly to be moved from a specific ; location in the fuel pool to a pre-assigned location ; in the core. It also provided the inal, ructions on : what control rods were to be exercised for : functional and sub-criticality checks for ! pre-defined core configurations. EC moves to be i ande during the fuel loading were also included. - Most of the changes required to the fuel loading i sheets during fuel loading were to move the ECs earlier due to high count rates experienced when ' fuel assemblies and/or the neutron sources wre too l close to the RCs. The only other change involved using Control Rod 10-27 (instead of 06-27) for a sub-criticality check due to an accumulator problem i with Rod 06-27.
- Four RCs (one per quadrant) were used to monitor (
the count rate from the start of fuel loading up to ; the point when 532 bundles were loaded in the core. ! In order to keep the E C count rate within a 1 desirable range and to accossodate an increasing ! core size, it was necessary to move the E Cs outward
- by approximately one cell routinely as fuel loading progressed. The location of E Cs was selected to
- ensure that each quadrant of the core was adequately monitored. (Sea Figure 3 3 18)
The upscale alara setpoint was set at 1 x 105 e andtheupscaletripsetpointwassetat2x10p ( cps for each RC. The downscale rod block setpoint j was 3 cps. The ECs were checked for flux response i either by control rod pulls during scheduled I st4 criticality checks or by lifting the E Cs [ partially out of the core. These flus response t checks were made at least once every eight hours [ during fuel loading and prior to the resumption of e fuel leading when fuel loading was delayed for eight i hours or more. In ad11 tion, the 51grval-to-Noise ratio was calculated for each R C prior to start of fuel load, during any required reverification of , plant initial conditions and every time the ECs j were moved to a new location. (See figure 3 3-2) ; I r - -_.___- - l Supplement 9 Pago 3 3-3 7 Four SRMs (one per quadrant) were used to monitor the neutron count rate starting from the point when 532 bundles were loaded in the core to the completion of fuel load (764 bundles). With the SRM detectors connected to the SRM instrument channels, therodblockandgi.supscalegripsetpointswere set down to 1 x 10 and 2 x 10 respectively, since no previous saturation test was performed on the SRM detectors. The down scale rod block setpoint was 3 cps. The SRM flux response check was performed at least once every eight hours during the fugl loading process by partially withdrawing aach SRM. Fuel loading commenced on March 20, 1085 with the loading of four fuel assemblies around the central neutron source. The loading continued in control cell units that sequentially completed each face of : an increasing square core loading in a clockwise l direction until a 12 x 12 square was completed with synnetry about the center source. The thirteen control cells (52 bundles) needed to form a 14 x 14 square array of bundles around the center Control Rod (30-31) were loaded next. The remaining cont;ol cells were loaded, one on each face at a time, ir a clockwise manner, such t .t the core was rotationally synaetric after every four control (See Figure 3 3-3) i cells had been loaded. Control rod functional and sub-criticality checks were performed either after every cell (first 4 cells in the core), or after every two or four cells as dictated by the detailed fuel loading sheets. The purpose of the sub-criticality checks was to ensure that it was safe to load the next control : cell (s). ; For each bundle a visual verification was performed to ensure tha; the bundle was proparly grappled before the bundle was lifted from the fuel pool racks, that there was adequate clearance on all - sides while the bundle was be*ng moved to the . reactor cavity and that it wes loaded in the core in ! the proper location with the proper orientation. Also, physical verification was made of the fact that the bundle was ungrappled before the hoist was raised. Similar verifications were sade for the blade guides lifted out of the core and the FLC moves made during the fuel loading process. 1 1 Supplement 9 Page 3 3-4 A 4 7-by-day account of the fuel load progress is given in Figure 3 3-5. Most of the problems that caused delays were related to the refueling bridge , (limit switch, power loss, grapple indication, air ' hose break, etc.). Fuel loading was halted on Sundays in order to perform required weekly surveillances on FLC/SRMs, IRMs, APRMs and the refueling bridge. During the fuel loading process, FLC/SRM count rates were monitored periodically and 1/M calculations were performed and plotted for each FLC/SRM and for the average of the four FLC/5RMs (See Figure 3 3-6). The average 1/M plot was used to prvject the estimated number of bundles for criticality. If criticality was projected during the next loading increment then the increment size was reduced between 1/M calculations. Strong geometric effects were seen, particularly during the first few bundles loaded in the core and also whan the bundles were loaded c. ear and FLC. These geometric effects resv]ted in erronious (but highly conservative) projections which often resulted in very small increment sizes (1 - 2 bundles) between 1/M calculations. After eighty bundles were loaded in l' the core, the maximum increment size between 1/M calculations was reduced to one cell (4 bundles except for the peripheral locations where a maximum of five bundles were loaded between 1/M calculations). Bundle LJK 677 was identified to have a rusted char.nel fastener that had to be replaced. Some debris was identified in the core on bundles LKJ 398, LJK 506 and LJK 957. After fuel loading was completed, these bundles were pulied out of the core to correct the respective problems and reinserted back into the core. After the 12 x 12 square array of bundles was i completed, a p:rtisl core shutdown margin (SDN) demonstration was performed by withdrawing the i analytically determined strongest Rod (26 - 27) and a diagonally adjacent hod (22- 23) out of the core. l Sub-criticality with these two rods withdrawn demonstrated that there was at least a 0 38% delta K/K shutdown margin for the existirc core . configuration. Because the calculated Keff for the ; 12 x 12 array with the two rods withdrawn was : 0.9758, and the calculate < Kert for the full coae i Supplement 9 Page 303-5 I with only the strongest rod withdrawn is 0.97, sub-criticality for the partial core demonstratsd that the shutdown margin would be met throughout the i remaining fuel loading process. l The fuel loading was completed after fifteen days on April 4, 1985. All criteria were satisfied. l e 1 a f i t
- i l
l t i i I I l l .,_= Supplement 9 Page 3 3-6 - , FIGURE 3 3-1 NEUTn0N SOURCE IDCATION AND BLADE GUIDE ORIDftATION PRIOR TO FUEL IAADING y - ~ ~~ ~ ~ 71 l [ ] L 4 ifi 7f 7l iz 7 i&Mm _ __ t zt # 7F zt;fF 7' 7F & & 7( , l ~T Jf k 7 f M 7?" T 7sY $h k XX; /. f E W M zizt sfiM X M f M H M# ff f f f f ff n X /i l f-M ?# iF% d 74Mf M MM X X+ f M F E 79 M& M F M # d f AI E f f f 5%f dif M 7?- f M N K Xb ~ k 7 Y f k 7k 7F f7f 7F k 7h Yf / # M M # 7F M # 7%F # # 7ff M X if l ~~ [ h 7b 7h M7~ [ 7 j s le~ 7s 7 7 f D -# 4 zf f f d zf# #- r - l I ->giij;i,,iil 1 a f t f M Mf MR I i l 1 I i i I i i i i l i l I l l
- SOURCE (7) / BLADE GUIDE (185) i l
l l l l Supplement 9 Pag) 3 3-7 FIGURE 3 3-2 Signal to Moise Measurement DATE A B C D # OF BUNDLES (TIE) DETECTOR CPS S/M CPS 3/N CPS 3/N CPS 3/N IAADED 03-20-85 FLC to 24 to 99 10 32 3 10 24 Prior to (2019) fuel load 03-21-85 FLC# 50 49 60 59 50 49 80 79 4 (0005) 03-22-85 FLC# 50 249 50 99 60 149 70 174 48 (0340) 03-22-85 FLC# 6.8 16 38 9.8 6.5 64 6.0 5 96 (2005) 03-22-85 FLC# - - 70 34 - - - - 96 (22?7) 03-23-85 FLC# 5 4 12 11 - - - - 144 (2110) 03-25-85 FLC to 19 0 11 14.7 12 19 0 12 14.0 156 (1420) 03-26-85 FLC# 10 49 0 20 89.9 - - - - 196 (0020) 03-26-85 FLC# 38 189 32 159 40 159 4.8 15 260 (1915) 03-28-85 FLC# 30 99 4 39 35 116 2.5 73 388 (1116) 03-29 85 FLC 300 999 100 999 150 374 90 299 440 (0907) 04-01-85 SRM 16 159 12 119 40 399 15 149 532 (1528) *S/N Ratios obtained during FLC moves -FLC not moved i l l Supplercnt 9 Pcge 3 3-8 FIGURE 3 3-3 CORE 1/MDING SEQUENCE ' 1 N .,,es g n pd4,, 4,, ,. ar.c ,A,, ,,a,A,,, iow dr., ..A. ,a
- 1. . r,,
n .., ,.,,, s. .. .m u.t u ,
- m. 4,,,. a :C du. ',od,rA. .., %, ,, ad,u d., .d.,..p ,,
n ,,g u . n d.,,u.g. . .p, ,.g... yt .g,, , .A, g.o ,e., .,c, 1l p- w.,,uy,npm ., . ~ z ,, >> ,> .4... ., u ,, , 4,. 3.. ,. u. g,,., . ,.,, ' d*.m.L, e . . d.. ede ,o u , A> =d.mA.. ,,- - , u ., d,,,,,w,a.. m. - .. a , % . , , , , as .r ,. ,, .s y,k - ,,,,, uc , , , , , . , , , , , , , , , . , . , , , ,, , , , , , , _4 .. , , , /sTd,- .,,, &2 'u fJ
- e, f . G i s ndf U M e i 4, I,, 11 i#7 T' W4
- if,f 41. .i w n z.4. ,,, .,hu . , , . '4 O,,r, 7, , ,
O, ,,+s, ,e, ,d,, .m& c, r e,sms.g,c 3 j,u.u. rg 3, Om . itr'a: s 5,, , , ,,.,, 1 s.u, w.> a G, ,.% 3c 7u r we e' 69 v6 : A A 'st it et 30 6 td.4 .d. ,. cDd.Y itn., , . d> ,>4vs 31 A'su #a7s. 1s d;n, ,n1i. m":+nn. ,4, 4t6,.,c J ' a 6* %%W" !.4 M D:- #7 9.. ,'a4,. Ay Am.A.. A A fi f I.f 86, }} 1 A'a. '.? ) d.3 g )g,+6$ dk.a r . .-6' /J*!oAkrl '- . 4g4 7 g n y.. . . ,g,, y 2....> '.7e u:4., iy 2g;,9, , .y,,, ,p,,v.g,i.p. - p,,, ,,g , , , , ii it ic 3 . . ..
- ,p~,c,i =
- .q.d:,.: 3.. ,e,4 di rd. An >4> , A
- n. u s 1
. ,,, A, i,. e, .,.c,u,A> . , >'.- 6: >>,,, u- 4,. nhu 20 i l HI t : p,~-.>. 4,4.. -.g.m.'y.. a g... qn . ng, 4 . s-2.:4. -,. ,g.,r 24..e , ,4, ,a . 2im. ,.79, ,,. g..o . y.4 m at u . . * , nhm*,,, , s .A. .. .A. .n.,2 .A.mA, ,,rvA, e, . .i . A ,.* , . _.h,u. ., .t,y n io y m*..Am.d.. .- . . . . ~. , , . , 3,, .. .,. . ,s. .A, . .e .m4., ,A. , , 4, .y., U m $.E .T ME ;E E .w 1. ,, , , . . , W .s, .- .,, , .,.,,
- 5. E Y 3-
,,, 7 > . , 02 'uTv, -J.r: , , n. i;, , ,n :<T, ,, GT,, 6;, .a l t aI - - - = - } r + + + ++ + + , U, 3+ + -l- + +4 ++ + - ll ~ se E, ,n, FF + Y+i -l- -F +el- ++ + + F- I',
- r t + + tt+ + d'+ ++ -t- + bL 8
+ !+ + + + + , + + g# ++ + + + + !;_ ! + '+ + + J+ + -e -e a+ +~+; + + + + "< + i/ T T t + + f _+ + ++ + + + . ll 2l,- 1 : 3 P +; + t i + + ?+ # %tl# 4Ad- H " + +. + + # =Shg g + + '+ 1-t- ++ ll ll5 ' + -t- + + + + + + /e + + ++ + + ll 2 + + + + r + + -F + ++ +1 + '+i+' ."Li ' i+ + ++ + + -P + + + + + + LJ ;; E ++ + + + -# + + + + + H " a , 7+ + + + 4iJ + + + +, - , a ~C + + + !# + + +P ll u n=a: eJr= ara = = ren :a a s a a ++..,c - _ _ _ _ . . - - _ _ _ _ . _ , , - ,m -_ . - - Supplement 9 Pag 3 3 3-10 w FIGURE 3 3-5 Daily Fuel Loadire Progress BUNi)M S IAADdD Deft DAY TO DATE 00tWENTS 03-20-85 4 4 Fuel load started at 2130. 03-21-85 32 36 Rod Block limit switch malfunction. 03-22-85 62 98 03-23-85 58 156 03-24-85 0 156 Weekly surveillance on SRMs, IRMs, APRMs and Refueling Bridge. 03-25-85 38 196 Fuel load resumed at 1500. 03-26-85 82 278 03-27-85 84 362 03-28-85 76 438 03-29-85 66 504 Transformer #64 lost due to initiation of its deluge (fire protection) system. 03-30-85 28 532 0400 refuel bridge power cable problea. Cable cut and re-termed tu restore the system.
- G3-31-85 0 532 Weekly surveillance. FLC to SRM switchover.
04-01-85 14 546 Fuel load resumed at 2000. 4 04-02-85 74 620 C4-03-85 48 668 Air hose damaged when stuck center section of the mast was released and dropped. 04-04-85 96 764 Fuel load completed at 2350. Supplement 9 Pago 303-11 FIGURE 3 3-6 NUMBER OF BUNDLES LOADED 1/M Plot l-4 ^_ _- . : cs ,e 2. _ __ =m +ne r w65JW= F #5 38HM3 W45 ; m_ '. . a Mg@m%:p;;%r " L % ypNg% . W MI M5 u 1 . .lu y;m- =.' m+- ^1:--i hi :_e e m w 41- -w iwmm -= .===' - '~=m- a=. 3 f === ; n= K ---?M;n=2 -m i fa :"+t-ihh Rim s= .=_ s e= e ,=__=== . _ . .. -9 . _ _ , __-1 v n i, u rA+ . - -.~ u g ..- - .m - tyli n,*H . e y/ M
- - m% wm FA q+.c
+ y r; ^f*1- $5h dNUNd _ M d 32 5tt%*M; - 1 i
- wdnt ne
-t-f* h Wewd r@n 49 wM + <
- dt "~
Mew \s.% %*. Mg ,. Ax i:w c-a c- e n:=wbw:n = a- =a m : =.u= n -# :. * -r + w _ ~.; 5_ _ _ _..- . -N *$ 5h ! Ih:2'b b7i -$ N . k-- N M r. '.~15 5 N - IM'-i' __. _ _ , ___. __+_ _ _ _ _ -_ . _ _ ~ _ % M vara _. - =, +rwuw m- rz me-y ' r. - e < g [ ~ ,, 5 + T-. m.w =w= Wa -t!^p- rFM?rk ; .sn%<g .hY'NW pm T W.7., % n* tW f tH. j @s-ht(W ne b A TJ' nH, n-a ~--n a
- v. .-
. yi*
- n. .w na .m w.n
. 'y _t.-TS '77- 11Q Vl i.*'25 -.i E r '3 27;'l M : % d 4 E 2.::34 ONW WNil =~;$'D:1 --55d-E "-5-!:5% 55_% . _= E . __ -- c- 3._ . ---_ % ng4 - T,_El_. =t:4r _---._ c2'_m,._r ;"5 M i:2 bC. .__ . ._. WA/__, . . e
- t. ,w % -
b . j,'. y+ '** ' . - . "M'M *,""'"I~- m ,- t h 'i M i g -- A ,. . , _ - , *d h"'Tp' 'D K, 7 Wh s .4 8t"il 1-76m tim J gs ' *4,_ 9 4 -._- ._ J +9"_ 7 W _ _ _ y r Q U7 $1 N,CfMz5 LE g%g M3 il MHw # M thti P n%t;n%M WM fiHM tJO =L. . r #b -Mi: =li-=.-M #FE.irC-L=; M 5-1 p- .._ .._r-- --9------- ----r - - - , . ; _.=-_; --;- 1u -- . =:r -- ~~-':-': b MM ({, '_k .
- 7. yw ,4 w ,
, a -,4 ,1 -mJ e44 w M -, , ,4, p. -+r p . s gi' Mr. 7 . 4' n- IITT n..a w , ' y .i_.,_J.-1, ,~ ,. 4 {dTT Q _r+-.gw *}n---- -4 r-4+ -- g,.. . 4 ,e4 {: L.t, 'g'4__ ~2 W "t ".IT{ w -dt is .- . + . N< . ..I - ,- , e I. .m.twn+ .g g. . , 14 % .. .L< mer -n-
- g. . .. ,
. d : i+ t#e-m ed ,%p,_ gN mrm.W t . u e -,_ --.- - . 2. I
- 3
~ hII$:,_k1JI $b. ~ ~ ~ ~ < b ~ .Nb- 1:*_.
- 1* * $ **
=. 1 = -F._-a..._ =_ i,i_?_ _=. __._iri E.=J : i.=5;+5, .I -]Esl:;".ri t .e s see An *** &oc e se h* kUMBER OF BUNDLES LOADED Supplement 9 Pag] 3 3-12 FIGURE 3 3-7 (Page 1 of 2) FUEL IDCATIO11 YERIPICATIOli l s.t 8: t' 9 ej i; y si y v : .. vii. 4 ...,ei- i i, i. - gMRMd i ! !..
- 4 MM J M @ i. i
'5tRM M M Ms! 4 bt M WR M M d -4 W MMM RM M R M MMM MMW M R MMM MM Ri 14R RE: MMMR: d R WM MRMR: ":it M Mft M@K4:R M M MI+gMMM R >:{f4 R MMM MS 'd.iRF;MMMMS M RM M t ,J .g. Lp w.u.M,s.u Supplement 9 P:ge 3 3-13 FIGURE 3 3-7 (Page 2 of 2) FUEL I4 CATION YERIFICATION , EM N%l l%.:- " 1M g M M y,- ~ aM M M8: hor erM M M M M M D- !M M M M MMMW-
- MRM M Mt* bt>
lMMM M M14 Mb" iM M M M MM M P . l M R M M MM M &;- R RMM MM M t*- 14 RM t*MM M P ?MMM P MM& ! %MM EA & MF- , !!BM R MBk i i + .;,,+ i +PViWM ,e :.+ &' -: Supplement 9 Page 3 4-1 34 Full Core Shutdown Margin 3 4.1 Purpose The purpose of this test is to assure that the reactor will be subcritical throughout the first cycle with any single control red fully withdrawn and all other rods fully inserted with the core in its maximum reactivity state. I 3 4.2 criteria 1.evel 1 The shutdown margin of the fully loahd core with the analytically determined strongest rod withdrawn I must be at least 0 38 percent delta k/k plus R (an additional margin for exposure) where R 05 percent delta k/k. Level 2 Criticality should occur within + 1.0 percent delta k/k of the predicted critical. l 343 nesults The fully loaded core was made critical by withdrawing control rods following the B sequence, using the Reduced Notch Worth Procedure. This sequence contained the analytically strongest Rod ( 06-39, which was fully withdrawn before reaching criticality. Prior to performing the shutdown margin demonstration, as required by Technical Specifications, the shorting links were removed to put the Reactor Protection Systea in the non-coincidence scran mode. The point of criticality was demonstrated by withdra.#ing control rods following the order given in the rod pull sheets until an (approximate) 300 second period was observed with Group 3 Rod 18-51 withdrawn to notch Position 08. Moderator temperature was recorded at 960F. 1.ater, with muderator temperature still at 960 F, the reactor was then made supercritical by withdrawing Control Rod 10-43 to Position 08. SRM A,B,C and D 3easurements were taken every 30 seconds for three and one half minutes. Period analysir was performed by fitting the data linearly on a semi-log plot and Supplement 9 P:33 3 4-2 seasuring time to increase one decade from which period was calculated. The average period was determined to be 76.5 seconds. 1hgshutdownmarginofthefullyloadedcoreat 68 F with the analytically strongest rod withdrawn was determined to be 2.72% delta k/k. Level 1 criteria were satisfied since the measured shutdown margin was larger than R + 0 38% = 0.88% delta k/k where R is defined here as the analytical difference in shutdown margin (cold) at the most limiting point in the cycle and Beginning of Life - of the core. The difference in keff between the theoretical critical configuration and the actual measured critical configuration was found to be 0.28% delta k/k. This satisfies Lev d 2 criteria since criticality occured within 15 delta k/k of the theoretical critical eigenvalue. Supplement 9 Pag 3 3 5-1 35 Control Rod Drive System 3 5.1 Purpose Each control rod drive (CRD) was tested to measure insert / withdraw and scram times and friction dP levels in the CRD hydraulio system. This was done , to demonstrate that the CRD system operates properly over the full range of primary coolant temperatures and pressures. 3 5.2 criteria Level 1 Each CRD aust have a normal withdrawal speed less
- than or equal to 3 6 inches per second, indicated by a full 12 foot stroke in greater than or equal to 40 seconds. l The mean scraa time of all the operable CRD's with functioning accumulators must not exceed the following times (scram time is measured from the time the pilot scram valve solenoids are de-energized). i l
Position Inserted i From Fully Withdrawn Scras Time (Seconds) 46 0 358 i 36 1.096 t 26 1.860 6 3 419 l The mean scram time of the three fastest CRD's in a ; two-by-two array must not exceed the following times , (scram time is sensured from the time the pilot j scran valve solenoids are de-energized). i Posit'4n Inserted ! From Fu_l1y Withdrawn Scram Time (Seconds) 46 0 379 ! 36 1.161 ! 26 1.971 i 6 3 642 i l i. .~ - . . - - - . - - - , - - - _ _ . . _ . _ , _ . . _ _ _ Supplopent 9 (' Page 305-2 i Level 2 i gach CRD aust have a normal insertion or withdrawal ! speed of 3 0 (+ 0.6) inches per second indicated.by ! a full 12 foot stroke in 40 to 60 seconds. - If the differentini pressurc variation exceeds 15 ! psid for a continuous drive-!n, a settling test must ; be yrformed. In this case the differential i settling pressure should not be less than 30 ps11, i nor should it vary by more than 10 paid over a full l stroke, j 353 Results l l Insert / withdraw timing, friction testing, and scraa ' timing were performed on the CRDe at the conditsons specified in Figure 3 5-1. All of the individual control rods were scraa. time tested. friction tested and insert / withdraw timed durJng the Open Vessel test condition. Adjustments to so.m CRDs had to be done in some cases to bring insert / withdraw times into acceptance limits. During the friction testing, no pressure I differer.tial sensurements exceeded the criteria of 15 psid and no settling tests had to be performed. The four slowest rods in each s6quence w9re also scranned at reduced accumulator pressure. All test criteria were satisfied. During Heatup, the four slowest rods in each sequente were scraa timed at 600 psig and at 800 psig. Upon reaching rated toeperature and petssure conditions, all CRDs wer. scram tlaed. The eight slowest rods determind during Open Vessel and Heatup testing were then insert / withdraw timed, friction tested, and scrammed at reduceJ accoulator l pressure. Figare 3.5 2 shows the average scram time i of the eldht slowest rods, four in each sequence, at various reactor pressures empared to the maximum permissible. 1 1 } 1 Supplement 9 P:se 3 5-3 The specific results from our rated pressure testing are as follows: l Mean Scran Times l l Rod Position 1 46 I 36 l 26 1 06 l l Mean Scras Time for all l 0 302 1 0.852 l 1 398 1 2 501 1 1 80 sea. B rods (sec) l l l l _! l Mean Scras Time for all l 0.288 1 0.802 l 1 340 1 2.436 I I 97 Seq. A rods (sec) l I l l l l Mean Scras Time for AIL l0.295 1 0.826 1 1 368 l 2.1 67 l l rods, Seq. A and Seq. 3 (sec) l l 1 l 1 l (core averase) l l I l l l Mean Scram Time of the l 0 325 I 0 900 1 1.481 1 2.655 I l 3 fastest CRDs in a two-by-twol l l l l l array for All roda, Seq. A andl l l l ! I Seq. B (core average) l l l l l l In conjunction with the planned scran for the Shutdown from Outside the Control Room test performed in Test Condition One, lhe scran times for the four (4) slowest Sequence "A" control rods were detersined. All the scran times were within the acceptance criteria. Supplement 9 Paa3 3 5-4 FIGURE 4 5-1 CONTROL-ROD '*1VE SYSTEM TESTS Reactor Pressure with Core Loaded Test Accumulator Preop psig Description Pressure Tests 0 600 800 rated ~ ~ Position All All Indication Normal Stroke Times All All 4(a) Insert /Nithdraw Coupling All All Friction All 4(a) Scran Normal All All 4(a) 4(a) All Scram Minimus 1(a) 4(a) Scraa Zero 4(a) Scram (scrandischarg{') Normal volume high level) Scran Normal 4(b)
- a. Refers to four CRDs selected for contiruous monitoring based on slow normal accumulator pressure scran times, or unusual operating characteristics, at zero reactor presssure. The four selected CRDs aust be compatible with rod worth minimizer, RSCS systems, and CRD sequence requirements.
- b. Scras times of the four slowest CRDs will be determined at Test Conditions 1 and 6 during planned reactor scrams.
- c. The scran discharge volume fill time will be determined at Test Conditions I snd 6 during p1Lnned reactor scrans.
Note: Single CRD scrans should be performed with the charging valve closed (do not ride the charging pump head). l l Supplement 9 Page 305-5 r! CURE 3 5-2 Scras festing Results . L..I s _ .. . .i _ . )l .. .. . . 1 o. M. .aL.i hub !. .l- { ~ i ! < 4 '.~ } ' . l ..i!., ...._...d,.,.. ~ .. gg,,ende. ' ., I_ , . . l . 1. .. 1<. . .. .. Lm C I - i ! l ,~ ~i9WWWtsvittr---* 4 - :- } I . ,. - .... !.i._. p . L, ; l i q ._..} . . y ,t . ..q 6... .J
- i. i
, t.,- -" L j i ! i j v .u a ! . _ . ..:. I j : -- ! .[~ ] :! 9l...j . I f.-~- w. :-LL .:.. n.: .. I :.... . i I- M f ",.'.= .j c - . ,; . .y t . g _. 7 {'. _. - .. 7 - . . r, - _ a = r-n. { ., .a , } i . . .. . .. ._ .a . . q. . . _ .._.' . .. . .-x..... .. _-,_K ;. ..L , .t , _ 4 i 3 . . _ , m .. . . . . . . . . .'. ~ . . , q ,.- .- , . , rmA __* . ;- , .- ; y .( m. , , ~ i . } - ; -' 2 d li , ,~ t- ' . . . i. ---- = 3, . . v i i Y < a i -a . _Q- 10..:.a._M b0. . -- Et 3 00 _ . X_ .- i .' - t inertion . : . i .- i ' . . . . e , . . a- y- - _.; q. - . > ,g .: - .2 n.s . . g. . ._4._. , ; - }"-.. Average 901 serse'T1mes at varlow?resoutes, ;. [j.. T-L e . +: . . _ .. . . m elaue: . - - - - - - ,- - hrainnb _- _. , -. y .-. . _y , . . .. 4,, .. . i .. ,- uj.x { ._, . , - - e J .. ; 2 , .} . -: 1 . > ! ., -i c l A. . Actul i e , _.___t.._ ._ . _ . _ _ - . _ . . - . - 4 o l 1 -l - _ _1 i ._.,.__M...__ [ ;. __ .^* f ,I_ . . _ . . _ _ . _ . _ _ . _ ] '}, i ' 'l . i i __ i ._ _ __7_ _ ,t _ __ _ . . , . ._ _ _p _. . . _ . . , . _.. .. . _.. i l e n _..i_._4_ i .0e i 25.0 43 6 _, .N. o Sbo . .-lob .o . ?...- statten htsant., ige W _
- t .
. . . . . . , .....~4 * - - . * > ~ l , 9 f .t. nr - Supplement 9 Pago 3 6-1 36 Source We Monitor Performance and Control Rod Sequence Exchange 3 6.1 Purg.ose The purpose of this test was to demonstrate that the operational sources, source range monitor (SM) instrumentation, and rod withdrawal sequences provide adequate information to achieve criticality and increase power in a safe and efficient manner. The effect of typical rod movements on reactor power was also determined. 3 6.2 criteria l i Level 1 There must be a neutron signal count-to-noise count ratio of at least 2:1 on the required operable SMs. There must be a alniaun count rate as defined by Technical Specification on the required operable SMs. Level 2 None j 363 nesults Prior to the initial criticality in sequence B. the count-to-noise ratio for SM ( A, B, C and D) were 43, 149, 199 and 49 respectively. These ratios were well above the Level 1 criteria of 2:1. The minimum counts on the SMs ( A, B, C ar.d D) wero 20, 15, 40 and 15 cps respectively. These were well above the minimus Level 1 criteria required of 0.7 cps. SM readin;;s were also taken periodically during initial criticality in both sequences and IM readings were obtained during the initial heatup in sequence B. All test criteria were satisfied. Performance data was gathered during power ascension to 20% in Control Rod Sciuence A and Sequence B. At the end of each rod worth minimizer group, APM, feed flow, and steram flow values were recorded. - J Supplement 9 Page 3 7-1 37 isater level Mansurement 371 Purpose The purpose of this test is to measure the reference leg temperature and recalibrate the instruments if the measured temperature is different from the value assumed during the initial calibration. 31.2 Cr1Loria Level 1 Hone Level 2 The difference between the actual reference leg temperature (s) Lad the value(s) assumed during initial calibration shall be less than that amount that will result in a scale endpoint error of 1 percent of the fnstrument span for each range. 373 nasults Testing of tha level instrumentation accuracy showed thac scale end point errors when actual drywell temperatures and assumed calibration temperatures were compared were 0.708% 0.5545,1.0507% and 0 320$ for wide range (Div. I), wide range (Div. 11), narrow range (Div. I) and narrow range (Div. 11), respectively. The slight Level 2 criteria violation for Div. I narrow range level instrumentation was found acceptable following an evaluation performed by General Electric. It was previously intended to repeat this test to obtain another set of data with all the drywell coolers in operation. However, based on an evaluation performed by General Electric, the test results are acceptable and no further testing is required. 4 Supplc2:nt 9 , P;ge 3.8-1 38 IRM Performance 3 8.1 Purpose .'he purpose of this test is to adjust the intermediate range monitor system to obtain an , optimum overlap with the SRM and APRM systems. 3.8.2 criteria 1 Level 1 Each IRM channel must be on scale before the SRMs exceed their rod block setpoint. Each APRM must be on scale before the IRi:s exceed their red block setpoint. Level 2 None 383 Results During the initial criticality, all IRMs except IRM - D ghowed response prior to the SRM's reaching 5 x 10 cps. IRM D was repaired and tested satisfactorily at a later date. Range 6/7 overlap calibration was also co )leted for each IRM, except IRM G which was reading stratically. This IRM tas replaced and retested successfully. IRMs G and H u..astwent repairs during the outage that required retestinf of the range 6/7 ove lap. After some adjustments, overlap was again sucessfully demonstrated for both. All APRMs were shown to be anscale prior to any IRM ! exceeding its r:d block setpoint during a planh l shutdown in Test Candition One. it was notes that IRH channels C, E, F and H were ! not reading one-half decade below their range 9 rod block setpoints. Although Technical Speci.*tcation verification of overlap was satisfactorily performed in conjunction with Plant Surveillance procedures, t the test was reperformed after APRMs were adjusted at a higher power level. Results of this reperformance follow below. I. l Supplsr:nt 9 P;ga 3 8-2 l 1 IRM G Range 6/7 Overlap Calibration was reperformed I successfully during the plant restart in October ) 1987. This calibration was necessary due to the i replacement of the preamplifier for this IRM. During the startup of the rsactor after the Soring 1988 LLRT Outage, IRM/APRM overlap was checked and adjustments were made to the IRM gains to provide additional margin to the IRM high rod block setpoint. The Level I criteria that each APRM aust be on scale before the IRMs exceed their rod block setpoint (< 108/125 on range 10) was satisfied as is shown by the following final settings: APRM Reading IRH Reading All GAFs = 1.0 After Adjustment Range A 83 20/125 to B 79 13/125 to C 8.0 60/125 10 D 7.9 40/125 10 E 8.1 44/125 10 F 8.6 26/125 10 0 20/125 to H 29/125 10 Since the IPMs had been e.djusted, the SRM/IRM overlap was reverified during a reactor startup in August 1988, and it was shown that each IRM was on scale before the SRMs exceeded their rod block setpoint. l ' Cupplement 9 l I P2ga 3 9-1 39 LPRM Calibration 3 9.1 Purpose The purpose of this test is to verify LPRM response to flus changes and proper LPRM connection to neutron monitoring electronics and to calibrate the LPRM's to their calculated valves. , , 3 9.2 Oriteria > I4 vel 1 None Level 2 Each LPRM reading will be within 10 percent of its calculated value. 393 nasults The initial LPRM verification test was performed witile the Reactor was at rated pressure in the heatup test condition, in conjunction with scran time testing. Specific control rods were selected to be used for flux response checks based on their proximity to the LPRM strings. The withdrawal of these rods from Position 00 (FULL IN) to Position 48 (FULL OUT) was observed in terms of the LPRM flux response as the rod was withdrawn past each of the four LPRMs for the associated LPRM string. All 172 LPRMs (43 LPRH strings with 4 LPRMs per string) were observed, using Brush Recorders and STARTREC System for flux response. Initially, no flux respons: was observed on 25 of the 172 LPRMs For the LPRMs that showed flux response, the proper order of the LPRM response (D, C, B, A) was observed. During supplemental testing, it was found that sc.- LPRM detectors were connected in reverse order anc these were corrected. One detector was found damaged and had to be repaired. During Test Condition Or,c all remaining LPRMs were observed to show proper flux response following repair efforts. An initial LPRM calibration utilizing the Traversing In-Core Probe (TIP) System and the Backup Core t Limits Evaluation (BUCLE) program was conducted in Test Condition One. Utiliziug TIP traces, local LPRM readin6s, and heat balance information, a gain Supplement 9 , Page 3 9-2 1 1 adjustaent factor (CAF) was determined for each LPRM. These GAFs were then used to adjust the gains of the LPRMs and a followup test was performed to verify , criteria. Due te non-steady shte ccoditions, a j total of four full sets of IIP traces were made. i Upon completion of the test, a total of 23 LPRMs did i I not meet the above criteria. The majority of the failured were reasonably close to the criteria, or I were in the low power region of the cors where ) criteria can be ignored. During Test Condition Three relevant portions of REP 'i4.000.05, LPRM Calibration - Computer liet,ermination, were performed. This entailed f.erforming an OD-1 with a complete set of TIP traces, running a P1 to update the LPRM GAFs, obtaining an OD-10 Option 7 GAF edit, and obtaining the initial LPRM flux amplifier input currents. All 172 GAFs were reviewed, and it was determined that eight (8) GAF adjustments on the following LPRMs were necessary. 16-33A 48-17A 48-33A 24-25A 16-57A 08-17D 16-09A 32-49D These eight GAF values were outside of the 0.95 to 1.05 range, and were used to calculate new LPRM flux . amplifier input currents. Following these eight (8) LPRM GAF adjustments, an OD-1 with TIP traces was performed, a P1 ras run and an OD-10. Option 7 GAP edit was obtained. , Upon review of the GAF edit only one LPRM GAF was , outside of the 1.00 + 0.10 required range. LPRM , 32-49D was reading 070, and was diagnosed as a , drifter on the latest P1 edit. IGAF was manually set, a P1 was run, and the LPRM 32-49D had a GAF of 1.0. Upon completion of REP 54.000.05, a n 172 LPRM readings were verified to be withir. 10 percent of their calculated readings, thu= satisfying the Level 2 criteria. J Supples;nt 9 P2ge 3 9-3 During Test Cor.dition 6, relevant portions of REP 54.000.05, LPRM Calibration - Computer Determination, were performed. This entailed performing an OD-1 with a complete set of TIP traces, rt Ing a P1 to update the LPRM GAFs, obtaining ai. OD-10 Option 7 GAF edit, and obtaining the initial LPRM flux amplifier input currents. All 172 GAFs were reviewed, and it was determined that 52 GAF adjustner.ts were necessary. LPRM 40-17A is failed low and is bypassed. These 52 values we:e outside of the 0 95 to 1.05 range, and were tv,: J to calculate new LPRM flux amplifier input t-trents. Following these $2 LPRM GAF adjustments, an OD-1 with TIP traces was performed, a P1 was run and an OD-10, Option 7 GAF edit was obtained. Upon review of the GAF edit, all but three LPRM GAFs were between .91 - 1.11, the required range, thus satisfying the Level 2 criteria. LPRM 40-17A is failed low and bypassed, therefore has a GAF of 0.0. LPRM 08-17A has a GAF of .91 and 24-33D has a GAF of 1.11. Since these GAFs were equal to the criteria value and not outside of the tolerance, no further adjustments were made based on that and the location of these LPRMs not being near critical fuel segments. This concludes the planned LPRM Calibratiens during the Startup Test Program. l 1 Supples;nt 9 l P2ga 3 10-1 1 3 10 Average Power Range Monitor Calibration 3 10.1 Purpose The purpose of this test is to calibrate the APRM system. 3 10.2 criteria Level 1 l In the startup mode, all APRM channels must produce a scram at less than or equal to 15 percent of rated thermal power. The APRM channels must b3 calibrated to read equal to, or greater than the actual core thermal power. Recalibration of the APRM system is not necessary from a safety standpoint if at least two APRM channels per RPS trip circuit have readings greater than or equal to core power. Technical 3pecification and fuel warranty limits on APRM scram and rod block shall not be exceeded. Level 0 If the above criteria are satisfied, then the APRM channels will be considered to be reading accurately if they agree with tha heat balance to within (+7, -0) percent of rated power. 3 10 3 Results During heetup, each APRM channel was calibrated to read greater than or equal to a manual calculation of Core Thermal Power based upon a constant heatup rate analysis. The APRM scram trip setpoints were also adjusted to produce a scram at less than 15% of rated power. The Level I criteria was satisfied. An initial APhM calibratiou was performed during Test Condition One at a Reactor Power of 13 3%. All APRMs were adjusted to read within (+3, -0)5 of calculated core thermal power, as determined by a manual heat balance calculation. A second APRM calibration was performed later in Test Condition One when core thermal power (CTP) was determined to be 15.56% as de' ermined from a manual heat bslance calculation. APRM gain adjustments were then evaluatud and the APRMs adjusted to read 16.0% which is +0.44% above CTP and satisfies the above Level 2 4 criteria.- Supp10c;nt 9 Pzg3 3 10-2 During Test Condition two, following a full core LPRM calibration, each APRM channel was calibrated to a reactor power of 48.4%. This reactor core thermal power was ca?culated by heat balance, and the six APRMs were calibrated to read within (+7, -0)$ of the 48.4% power, thus satisfying Level 2 criteria. This also ensured that the Level 1 criteria requiring that the APRM channels be calibrated to read equal to, or greater than the actual core thermal power was set. Finally, the Scale Factor was determined to be equal to 1.0 since no APRM gain adjustaents were imposed. This satisfied the Level I criteria requiring that Technical Specifications and fuel warranty limits on APRM scram and rod block shall not be exceeded. During Test Condition Three, the Process Computer was used to determine a core thermal power of 48 3%. No APRM gain adjustments were imposed which allowed the Scale Factor to be set equal to 1.0. Therefore, the sir APRM desired readings were determined to be 48 3%. The six APRM readings taken locally at Relay Room Panel H11-P608 revealed that the absolute differences between the desired and current APRM readings were within (+2%, -0%) except for APRM B which initially read 48.2%. Therefore, APRM B was adjusted by changing the setting of the R16 gain potentiometer to read greater than 48 3% CTP. The final APRM readings at that power were as follows: APRM A 50.0 APRM D 49.2 APRM B 48.8 APRd E 48.6 APRM C 49 0 APRM F 49 2 The scale Factor was determined to be equal to 1.0 and all the APRMs are reading greater than core thermal power. This satisfied the Level 1 criteria. As seen by the data above, the Level 2 criteria is also satisfied. During Test Condition Five, the Process Computer was used to determine the core thermal power of 71.7%. No APRM gain adjustments were imposed which allowed the Scale Factor to be set equal to 1.0 c.nd, therefore, the desired APRM readings were determined to be 71.7%. The actual APRM readings taken locally at the Relay Room panel H11-P608 were between 70.0% I Supplemint 9 PIge 3 10-3 l to 71% of rated power. All six (6) APRMs were adjusted to read greater than 717% CTP by changing the setting of the R16 gain potentiometer. The final APRM readings at that power were as follows: APRM A 72.0% APRM D 72.0% APRM B 72.0% APRM E 72.0% APRM C 72.0% APRM F 72.0% Since the Scale Factor was determiaed to be equal to 1.0 and all APRMs are reading greater than core thermal power, the Level 1 criteria is satisfied. The Level 2 criteria is also satisfied by the above readings. During Test Condition 6, the Process Computer was used to determine the core thermal power of 96.8%. No APRM gain adjustments were imposed which allowed the Scale Factor to be set equal to 1.0, and therefore the desired APRM readings were determined to be 96.8%. The actual APRM readings taken locally at the Relay Room Panel H11-P608 were acceptable except APRM "E" which was adjusted from 96.0% to 97.0% by changing the setting of the R16 gain potentiometer. The final APRM readings at that power were as follows: APRM A 97.0% APRM D 97.5% APRM B 97.0% APRM E 97.0% APRM C 97.0% APRM F 97 5% Since the Scale Factor was determined to be 1.0 and all APRMs were reading greater than the calculated core thermal power of 96.8%, both the Level I and Level II criteria are satisfied. I i y - Supplem;nt 9 P:ga 3 11-1 3 11 Process Computer 3 11.1 Purpose The purpose of this te,L is to verify the performance of the process computer under plant operating conditions. 3 11.2 criteria i Level 1 None Level 2 Programs OD-1, P1, and OD-6 are considered operational when the MCPR, the maximus LHGR, the maximum APLHGR, and the LPRM gain adjustment factors calculated by BUCLE and the process computer agree with the tolerances specified in the FSAR. Remaining programs will be considered operational on the successful completion of the static and dynamic testing. 3 11 3 Results The TIP System consists of five identical probes used to measure and record ti.e axial neutron flux profile at 43 radial core locations. The recorded information is used by the Process Computer to calibrate the fixed in-core Local Power Range Monitors. Each probe is driven into and withdrawn from the core by its associated drive mechanism. In order to operate automatically, the TIP drive control units must be programmed with the probe position at top and bottom of the core. These top and bottom limits are programmed and verified in the TIP cold alignment. This portion of the test was performed successfully by hand-cranking the TIPS to the top of the core and setting the core limits based on the resulting position readings. In order to follow and read data from the TIP machines, the Process Computer must receive position information and flux signals from the TIP System. This interface is tested in the Static System Test Case by running the TIP machines in various ' configurations and verifying the proper responses on the Process Computer. N Supplen;nt 9 PJg2 3 11-2 The Ctatic Systen Test Case had two objectives: verification of the progran logic and checkout of the TIP interface. The first objective was successfully achieved, but the TIP interface checkout was unsuccessful due to a problem with the TIP System that resulted in the loss of TIP position indication. This original position indicatio1 . poblem was repaired. As part of the Test Condition One testing, the TIP top and botton core limits were reverified under hot conditions, and the TIP interface with the X-Y plotter was also verified to function properly. Following repairs to TIP "C" ball valve, a process computer interface problem, and TIP "B" Logic, a successful OD-1 was obtained from the process computeri It was noted that a three (3) second delay was occurring between X-Y plotter traces and the nachine normalized, full power adjusted TIP array. This probleia was corrected prior to the OD-1 portion of the Dynamic Systen Test Case. The Dynamic System Test Case was performed during steady state ccnditions with reactor power at approximabley 20%. The testing included:
- 1. Verification of the Computer Outage Recovery Monitor (CORM) to initialize necessary variables and exposure arrays as part of initial plant computer startup and to allow for controlled set of data in further system testing.
- 2. Verification that all required plant sensors for NSS programs are being properly scanned.
3 Verification of the heat balance subroutine used by OD-3 and P1 by comparing it with a manually calculated heat balance.
- 4. Performing an LPRM calibration to verify the l operation of OD-1 prior to the verification of I thermal liuit calculations.
L
- 5. Verification of thermal limits esiculations and core power distribution.
- 6. derification of the exposure updating programs P4 (10 Minute Core Energy Increment), P1 (Periodic Core Evaluation), P2 (Daily Core Performance Sunnary) and P3 (Monthly Core Performance Summary).
SJpplement 9 P2ge 3 11-3 7 Verifying key variable memory locations and performing manual calculations to verify the remaining NSS software at steady state operation and symmetric rod pattern. Thermal limit and LPRM calibration factor calculations were verified in conjunction with the DSTC. The verification was performed by taking the same data that is input to the P1 program, for its calculation, and inputting it into an approved offline computer program (Backup Core Limits Evaluation (BUCLE), which also performs the P1 calculations. The resulting thermal limits and LPRM calibration factors were verified against the criteria. In all instances the results were in the same fuel assembly and the results are as follows: Parameter Location P1 Results Bucle Results % Error Max LHGR 33-52-13 3 78 3 78 05 Max MAPLHGR 27-10-13 3 30 3 30 0% Min CPR 27-10 3 877 3 876 .02% P1 Result - Bucle Result % Error =
- 100%
P1 Result The Local Power Range Monitor (LPRM) gain adjustment factors calculated by BUCLE and the process computer were verified to agree within 2%. Programs OD-1, P1, 00-6 and the remaining NSS programs were considered operational upon the satisfactory performance of this procedure. During Test Condition Three, a Process Computer - BUCLE Comparison was performed at steady-state conditions at 48.4% reactor power and 93% core flow. With P1 blocked, the following list of process computer edits were obtained and compared to the respective BUCLE edits: RCAL GAF W PBUN EBUN NSS Core Perfcraance Log Thermal Data in Fuel Assembly IX, JY The 12 Bundles Closest to CPR Limits The 12 Highest Ratios of a Bundle MAPLHGR to its LIMLHGR Target Exposure and Power Data Supplement 9 P:ga 3 11-4 Each process computer value was verified to agree with each BUCLE value to within + 25 (FSAR tolerances). An MCPR of 2.819 was calculated by P1, and an MCPR of 2.821 was calculated by BUCLEs PINEWRP, each for bundle 17-18. These values are within 0.07% of each other, therefore satisfying the Level 2 criteria. An MLHCR of 5 76 was calculated by P1, and an NLHGR of 5.75 was calculated by BUCLEs PINEWRP, each for bundle 17-26-11. These values are within 0.17% of each other, therefore satisfying the Level 2 criteria. An MAPLHGR of 5.05 was calculated for bundle 17-26-11 by both P1 and BUCLEs PINEWRP. Therefore, the Level 2 criteria was satisfied. The process computer OD-10, Option 7 GAF edit was compcred to the BUCLEs EDITMAP GAF array. lhe values were verified to agree within ! 25, therefore satisfying the Level 2 criteria. During Test Condition Three, the Process Computer - Power Change Verification was performed to demonstrate the performance of the OD-4 and OD-5 programs during power changes. The test was performed in three sections where two of those sections dealt with the performance of OD-4 and OD-5 programs after a large power change (> 20% ; of rated power) from either recirculation flow alone or control rods alone, as compared to P1 program. One section dealt with comparison of symmetric and non-symmetric P1s. Symmetric P1 and Non-Symmetric P1 Comparison Once steady-state conditions were established, P1 program was run with the synaetry flag set to ' reflect airror symmetric conditions. After P1 run was completed, the Bundle Power array (PBUN) was obtained from OD-10 Option 22. Next, the syssetry flag was set to 3 (asymmetric) and P1 run again and MFLCPR and MFLPD were compared against the last P1 (syssetric) run. Also the Bundle Powers were compared against the last edit and all these parameters were observed to be within 5% ras. Once all the required edits were obtained, the core symmetry flag was restored to airror synaetry. Supples;nt 9 P;ga 3 11-5 Large Power Change (using flow only) This section was performed by establishing steady-state conditions, running OD-4s for at least 4 10 different rods, running OD-5, running P1 and then ) blocking P1. With P1 blocked, power was raised by l flow alone, at least 20% of rated pwer and once steady-state conditions were established at the new power level, OD-4s were run for the same rods t.s j before, OD-5 was run and then P1 was run. OD-4 and ! OD-5 edits were then compared against P1 before and l after the power change and the results were as I follows: OD-5 vs P1 OD-4 vs P1_ (% diff) (5 diff) Power (%) MFLCPR MFLPD McLCPR MFLPD 53 49 0 348 0.05 0,632 0.202 73 57 0.265 0.141 1.550 0.572 The overall power increase was 20.08% of rated power and the "SML NSS" video alarm flag was observed during power ramp. OD-18 Verification Data was completed and the core flow (WT) was verified to be between WLO and WHI. OD-19 Verification Data was comp]cted and the APRM A value was verified to be within the proper power band. The high and low powers pertaining to the power band from Data Class 15 were compared to tne OD-3 edit and were found to be within 1% of each other. g ge, Power Change (using control rods only) This portion of the test was performed after steady-state conditions Here established at 26.085 power and all required edits (OD-4s, OD-5 and P1) were obtained before blocking P1. Power was increased by rods alone to 48% power to provide an overall power change of 2192% of rated. Once steady-state conditions were reached, the required edits (OD-4s, OD-5 and P1) were obtained for i comparison. The results were as follows: l l OD-5 vs P1 OD-4 vs P1 (5 diff) (% diff) Power (%) yLCPR MFLPD MFLCPR FLPD 26.08 0.1 0.167 0.13 0.13 48.00 1.63 4.57 1.82 4.83 l Supplement 9 1 Page 3 11-6 I l At the conclusion of this segment of the test, P1 was restored to normal operation. Based on the above successful testing, 00-4 and OD-5 programs are considered operational, which satisfies the Level 2 criteria. During Test Condition Three, with the reactor operating at a steady-state power of approximately 49%, the Process Computer PCIOKR (Preconditioning Interia Operating Management Recomacndation) Verification was performed to verify the correct operation of the OD-11 program. There are fifteen options associated with the OD-11 program: Options 1 through 11, Options 66, 77, 68 and 99. OD-11 calculates and edits data pertinent to the monitoring and t.pplications of the PCIOMR. This program has six options (1 through 5 and 11) which edit information concerning the present power distribution and the stvred preconditioned envelope, twe options (6 and 7) concerning predicted power increases due to control rod withdrawals, four options (8, 77, 88 and 99) which permit monitoring of the preconditioning ramp rate on a model basis, and two options (9 and 10) which allow operators to establish and maintain the preconditioned envelope. In addition, Option 66 is available for automatic editing or suppression of the Option 3 and 6 edits. The verification testing was performed as follows: Data Interrogation The first step was to save PCIKON, KWTH, IEXPC and IPC arrays on magnetic tape. During the course of the test, severtl of the arrays were changed to force messages on the alarm typer or other edits which would facilitate checkout of the OD-11 software. At the end of the test, this ma;;netic tape was used to restore these arrays back to their original values. Since this test had originally been scheduled for Test Condition Five and was now being conducted at a ralatively low power level in Test Condition Three, the threshold nodal power was reduced to 5 kW/ft (from 14 kW/ft) so that there would be some nodes in the core that would exceed this artificially low l threshold va)ue. i i Nodal power edits on OD-11 Option 4 were checked against hand calculation for selected controlled and i Supplem;nt 9 22ga 3 11-7 uncontrolled nodes and were found to be in general , agreement. Nodal exposure and control rod positions I were satisfactorily compared against OD-10 edits. ; Nodal powers, pre-cor.ditioned power values, the envelope powe" values and exposure edits on OD-11 Options 1, 2, 3, 4 and 5 were compared for consistency. Envelope Updating OD-11 Option 1 was run to obtain the edit of Nodes versus DELTA-E intervals. OD-11 Option 9 was run for a selected non-zero DELTA-E interval and verified to properly updato the nodes in that interval. Next, OD-11 Option 2 was run to get an edit of P-PC versus number of nodes in each interval. Since the nodal power for all nodes was below the pre-set minist a pre-conditioned value, process computer parameter PF was increased fcr a selected node to provide a situation for which the nodal power exceeded its pre-conditioned value and would be a candidate for envelope updating. OD-11 Option 10 was run and this selected node was verified to be properly updated before PF was restored to its origins' value. Predictive Overpower Model OD-11 Option 6 was run to get control rod notch positions and rod withdrawal permissives then compared against OD-7 and OD-11 Option 5 and Option 7 for consistency. PCIKON (1) was lowered from 7.945 to 5.5 to create a situation where rod withdrawal would not be permitted on OD-11 Option ,
- 6. OD-11 Option 7 was run to verify that the predictive model was working properly before PCIKON (1) was restored to its orignal value.
Automatic Alars and Initiation This segment of the test involved checking out OD-11 Option 66 which turas on (or off) the OD-11 Option 3 and Option 6 edits. These options (3 and 6) only run if the fraction of feedwater finw is greater than PCIKON (3). This portion was checked by actually reducing PCIKON (3) to a value below the fraction of feedwater flow and either obtaining the Option 3 and 6 edits or verifying the edits were suopressed. The Feodwater Alarm setpoint calculated by the OD-11 program was verified against the hand calculations. Supple 2nt 9 P2ge 3 11-8 PCIKON (5) was changed from 0 3 to -10 to create a situation which would provide an overpower alara. Once this was verified, PCITON (5) was changed to 10 which produced a message that the overpower alarm had cleared before the original value of PCIKON (5) was restored. Control rods were aaneuvered to check out alarms associated with overpower situations due to rod pulls. PCIKON (4), and PCIKON (7) and WTFACT array was changed to create sitestions which would be identified by the OD-11 program as a potential 1 overpower condition resulting in an alarm and/or P1 initiation. The effect of KNOT variable was checked by setting it to .2 and observing that P1 aborted after initiation due to "asymmetry" as expected. It was also verified that for a significant change in Core Thermal power, OD-11 Option 3 and 6 edits will be printed after P1 even if these edits are turned off by Option 6o. Static Test of OD-11 Resp Monitor The first portion of the OD-11 Ramp Honitor check was to determine the proper value of threshold power, PCIKON (1), PCIKON (3) and IPC array which would enable the selection of five nodes representing margin to envelope (P - max (PC, KWTH)) in the following five segments: Lets than .055 kW/ft Greater than .055 kW/ft but less than 0.0 Between 0.0 and 0.2 kW/ft Slightly above 0.2 kW/ft Largest P - max (PC, KWTH) value Once these five nodes were selected, OD-11 Option 8 was initiated by running OD-11 Option 77 and OD-11 Option 88. The subsequent OD-11 Option 8 and Option 3 edits were compared for consistency. The parameters that were checked for consistency / accuracy were the nueber of overpower nodes, margin to envelope, peak nodal power, ramp rate, nodal exposure and that the PC value falls into the proper segment for ramping. Upon running another P1 it was verified that all the previously flagged overpower nodes were properly initialized. Of the five selected nodes, four of the nodes were observed to behave predictably but the P-PC value for the highest power node was somewhat lower than expected. The calculated P-PC was 1.2 which was outside the 1 32 2 055 kW/ft range. Upon further i investigation it was observed that the original l l i ,,,--n. - - - , - - - - - ~ . Supples;nt 9 Pige 3 11-9 setup to perform this portion of the test was based on the peak noda1 power of 6.66 kW/f t. However, after approximately five hours when this portion of the test was being performed, the maximum nodal power was 6.56 kW/ft which is 0.1 kW/ft lower than the original nodal power and acceptably explains this discrepancy. The Process Computer overpower alara setpoint based en feeditater flow was checked against hand calculations. The alarm setpoint was reduced to check the program that provides the "potential overpower alarm" prior to restoring the setpoint back to its original value. The pre-conditioned power values and the ramp rates were successfully checked for accuracy against hand calculations. The OD-11 Option 8 ramp monitoring program was observed to properly account for step change in nodal power due to OD-2 runs and the nodal PC values from the OD-11 Option 4 edits were successfully checked against hand calculations. Finally, the OD-11 Option 8 auto termination feature based on low ramp rate was verified to be properly functioning before the PCIKON, KWTH, IEXPC and IPC arrays were restored to their original values. Based on the above successful testing, the PCIOMR program, OD-11, is considered operational which satisfies the Level 2 criteria. During Test Condition 6, a Process Computer - BUCLE Comparison was performed at steady-state conditions at 96.8% reactor power and 97 7% core flow. With P1 blocked, the following list of process computer edits were obtained and compared to the respective BUCLE edits: RCAL CAF W PBUN EBUN NSS Core Performance Log Thersal Data in Fuel Assembly IX, JY The 12 Bundles Closest to CPR Limits The 12 Highest Ratios of a Bundle MAPLHGR to its LIKLHGR Target Exposure and Power Data Each process computer value was verified to agree with each BUCLE value to within + 2%. Supplemant 9 P ga 3 11-10 An MCPR of 1.483 was calculated by P1, ar.d an MCPR of 1.479 was calculated by BUCLEs PINEWRP, each for bundle 15-48. These values are within 0.27% of each other, therefore satisfying the Level 2 criteria. An MLHGR of 11.63 was calculated by P1, and an MLHGR of.11.67 was calculated by BUCLEs PINEWRP, each for bundle 43-46-4. These values are within 0 34% of each other, therefore satisfying the level 2 criteria. An MAPLHGR of 10.13 was calculated by P1, and an MAPLHGR of 10.12 was calculated by BUCLEs P1NEWRP, each for bundle 17-50-8. These values are within 0.09% of each other, therefore satisfying the Level 2 criteria. The process computer OD-10, Option 7 GAF edit was compared to the DUCLEs EDITMAP GAF array. The values were verified to agree within 2 2%, therefore satisfying the Level 2 crite.ria. The following discrepancies were noted: When comparing the value of Fuel Segaent Quality (QUAL), Segment Void Fraction (VF) and Sagaent Power (POW) for the incore limiting bundle u/Md for the comparison (43-46), the BUCLE value and the OD-6, Ontion 2 value differed by greater than 2% for several nodes. OD-6, Parameter Node Cpt. 2 BUCLE Scror POW 24 0.0735 0.072 2.045 QUAL 3 -0.0076 -0.0073 3 95% QUAL 4 0.0073 0.0077 5.20% VF 2 0.035 0.035 2.78% Since the computed values were very small, the small relative differences resulted in large percentage differences. Since the actual absolute differences are small, these discrepancies are not considered significant. When comparing the values from the LPRM RCAL Array, the BUCLE value atid the OD-10 value for LPRMs 16-57-D and 56-33-D differed by 2.57% and 2.195, respectively, thereby exceeding the requirement for agreement to within 25 The raw LPRM readings manually inputted into the BUCLE Prog?an were rounded values (18% vice 18.45% for 16-57-D and l l Supplen:nt 9 i P ge 3 11-11 l 23% vice 23 5% for 56-33-D). As a result, the BUCLE Program yielded RCAL values which exceeded the 25 requirement. Had the inputs been made using at least the first deciaal, the BUCLE RCAL Array values would have been very close to the 00-10 values (to much less than a 21 difference) and, therefore, this deviation is acceptable. l l This concludes the series of tests performed on the j Process Computer during the Startup Test Program. , l l 4 Supplen;nt 9 P2ge 3 12-1 l 3 12 acIC system 3 12.1 Purpose The purpose of this test is to verify the proper operation of the RCIC system over its expected operating pressure range. J 3 12.2 criteria Level 1 The average pump discharge flow must be equal to or greater than the 100-percent-rated value after 50 seconds have elapsed from initiation on all auto starts at any reactor pressure between 150 psig and rated. With pump discharge at any pressure between 250 psig and 100 psi above rated pressure, the I required flow is 600 gps. (The 100 psi is a l conservatively high value for line losses. The measured value may be used if available). The RCIC turbine shall not trip or isolate during auto or manual starts. Level 2 To provide a margin or. the overspeed trip and isolation, the first and subsequent speed peaks on the transient start shall not exceed the rated speed of the RCIC turbine by more than 5 percent. For small speed or flow changes in either manual or automatic mode, the decay ratio of each recorded RCIC system variable must be less than 0.25. The turbine gland seal condenser system shall be capable of preventing steam leakage to the atmosphere. l The delta P switch for the RCIC steam supply line I high-flow isolation trip shall be adjusted to actuate at 300 percent of the maximum requ!. red steady state flow, with the Reactor assumed to be near the pressure for main relief valve actuation. Supplcacnt 9 Pcge 3 12-2 3 12 3 Results During the Heatup Test Condition, the RCIC pump suction and discharge was lined-up in a closed loop with the condensate storage tank. 'lhe system was subjected to negative and positive 10% step changes in flow at system flows of 600 gpa and 270 gpa using both a step generator and the RCIC flow controller. l Minimum flow data was also taken at a speed of 2000 l rpm and a PCIC quickstart was performed. l The RCIC system was able to supply 600 gpa at a discharge pressure of 1140 psig in 35 seconds when automatically started using 940 psig steam from the vessel. The K72 time delay relay was set down from 10 see to 5 see to prevent the RCIC turbir.e from coasting down excessively before the opening of the Steam Admission Valve, thus reducing the experienced transient. The RCIC turbine did not isolate or trip during the auto and maunal starts. In addition, there were no RCIC turbine speed peaks or oscillations in RCIC systen variables in the transient testing. The RCIC system was also subjected to an extended run at rated flow conditions. RCIC performed satisfactorily with all system temperatures stabilized below alarm levels and a negative pressure maintained on the gland seal condenser system. All Level 1 and Level 2 criteria were satisfied except the RCIC steam supply high flow isolation trip setting. During the Outage for the replacement of the Main Steam Bypass Lines, engineering modifications to the instrument lines were completed that were expected to solve the problems found with the instrument sensing lines. l Upon recommencing Heatup in August of 1986, the RCIC EGM module was found malfunctioning and was replaced. Because of this and the instrument line j modifications discussed atove, the RCIC system was subjected to further testing including 10% positive and negative step changes in both speed and flow, and a quickstart. With the reactor pressure at 955 psig, the RCIC system was able to supply 600 gpm at a discharge pressure of 1143 psic in 33 seconds. All Level 1 and Level 2 criteria were satisfied except the turbin) gland seal system verification and the RCIC steam supply high flow isolation trip setting. Supplement 9 P:ga 3 12-3 Due to a failure of the RCIC Barometric Condenser Vacuum Pump, data did not show the existance of a vacuum on the vacuum tank as required by the test criteria. Subsequent work on the Barometric Condenser Pump corrected the problems and it was retested successfully. Data was also taken during this test to determine the actual 300% value for the RCIC steam supply line high flow isolation trip setpoint. However, the trip setpoints were not adjusted to these settings, but are being left at the current trip setpoints given in the Technical Specifications. The current settings as specified by the Technical Specification are set conservatively compared to the value calculated by the performance of this testing, yet provide ample margin to prevent spurious RCIC isolations on system automatic initiations. During Test Condition One, RCIC system testing consisted c,f a hot manual vessel injection, two (2) cold quick start vessel injections, a 150 psis CST to CST run, a 150 psig vessel injection, and a CST to CST run at rated pressure for baseline data. The only problem of any significance during any of taese runs was a turbine speed peak 29 rps above the Level 2 limit of 4725 rpm, which occurred during the initial hot manual vessel injection. Hinor adjustments were made to the RCIC control circuitry and the problem did not reoccur in subsequent tests. For the hot manual vessel injection, with the reactor supplying steam at a pressure of 915 psig, the RCIC pump delivered a flowrate of 3 600 gpm at a discharge pressure of 965 psig in 28.4 seconds. As discussed above, the turbine reached a maximum speed peak of 4764 rpm, which exceeded the Level 2 criteria. Based on data taken in conjunction with this test, it was determined that the actual line loss value for the RCIC system was 50 psid. For the first cold vessel injection, with the reactor supplying steam at a pressure of 918 psig, the RCIC pump delivered a flowrato of 3 600 gpm at a discharge pressure of 970 psig in 28.5 seconds. The maximum speed peak was 4686 rps for the RCIC turbine. Supplement 9 Pega 3 12-4 For the second cold vessel injection, with the reactor supplying steam at a pressure of 910 psig, the RCIC pump deliverad a flowrate of 3600 gpa at a discharge pressure of 970 psig in 29 2 seconds, with a maximum speed peak of 4488 rpm. During the 150 psig CST to CST run, with the reactor supplying steam at a pressure at 165 psig, the RCIC pump delivered a flowrate of 3 600 gpo at a discharge pressure of 271 psig in 22.0 seconds, with a maximum speed peak of 2818. During the rated reactor pressure CST to CST run, with the reactor supplying steam at a pressure of 920 psig, the RCIC pump delivered a flowrate of 3 600 sps at a discharge pressure of 1095 psig in 29 seconds, with no discernable speed peak as the turbine ramped up smoothly to a final speed of 4500 rpm. The 150 psig vessel injection was conducted with the reactor supplying steam at 160 psig. The system reached 3 600 gpa in an elapsed time of 21.5 seconds at a discharge pressure of 215 psig, with a maximum speed peak of 2641 rpu. RCIC testing was successfully completed with a 150 psig cold CST to CST baseline data test. With the reactor supplying steam at a pressure of 165 psig, the RCIC pump delivered a flowrate of 3 600 gpm et a discharge pressure of 360 psig in 19.5 seconds, with an initial speed peak of 141P, rpm followed by a smooth ramp to a final maximdm tpeed of 2766 rpm. Subsequent to the completion of testing, data gathered to determine the RCIC steam supply line high flow isolation trip setpoint was evaluated further by Nuclear Engineering. The results of this evaluation validate the flow equation used to determine the initial trip setpoint as presently listed in Tech. Spec. Table 3 3 2-2 and, therefore, no adjustment is necessary. This evaluation is detailed in Design Calculation #4595 Revision B. SupplcO: Int 9 x P ge 3 13-1 3 13 HPCI Systes NOTE: As discussed in memorandum NRC-87-0179, "Initial Test Program Changes", dated September 30, 1987, from B. R. Sylvia to U.S. Nuclear Regulatory Commission, Washington, D.C., the Level 1 criteria for system respor.se time to rated flow has been modified to agree with Plant Technical Specifications. 1he Level 2 criteria for margin to overspeed trip has been modified to reflect the control system hydraulic modifications which improved the stop and control valve response to a quick start. 3 13 1 Purpose The purpose Of this test is to verify proper operation of the High Pressure Coolant Injection (HrCI) system over its expected operating pressure range. 3 13 2 criteria 1.evel 1 The average pump discharge flow must be equal to or greater than the 100-percent-rated value with a system response time of less than or equal to 30 seconds as defined in Technical Specifications at any reactor pressure between 150 psig and rated. With put.p discharge at any pressure between 250 psig and 100 psi above rated pressure, the flow should be at least 5000 gpm. (The 100 psi is a conservatively high value for line losses. The measured value may . be used if available). The HPCI turbine shall not trip or isolato during auto or manual starts. Level 2 The turbine gland seal condenser system shall be capable of preventing steam leakage to the atmosphere. The delta P switch for the HPCI steam supply line high flow isolation trip shall be adjusted to actuate at 300 percent of the maximum required steady-state flow with the reactor assumed to be near main relief valve actuation pressure. Supplexn?, o Paga 3 13-2 ! l For small sped or flow changes in eithea manual or automatic mode, the decay ratio of each recorded HPCI systen variable must be less than 0.25 l The margin to avoid the overspeed trip shall be at l least 10% of the nominal overspeed trip setpoint of ! 5000 rpm, during all auto starts of the HPCI system. l 3 13 3 nesults Following setup of the control system, initial coupled turbine performance runs were performed on the HPCI system during initial heatup. Dynaalc stability chee's were conducted with the HPCI pump suction and d , charge lined-up in a closed loop to the ':ST. Flow step changes of ! 500 sps were introduced by the flow controller in automatic, with HPCI system flows at 5000 gpa and 2700 gpa. During automatic initiation testing of HPCI, a discharge flow of 5000 sps was reached in 23.4 seconds. Twenty-five seconds after the automatic initiation HPCI flow had reached 5310 spa at a discharge pressure of 1140 psig, 190 psig greater than reactor pressure. HPCI did not trip or isolate during any of the manual or automatic starts. Adequate margin was demonstrated on turbine speed peaks and oscillations of systen variables. An extended run was performed in which system temperatures stabilized at acceptable levels and the gland seal system performed satisfactorily. All Level 1 and Level 2 criteria were satisfied except for the steam supply isolation trip setpoint. During the extended Outage which started in the Fall of 1985, engineering modifications were completed that were expected to correct the problems experienced with the instrument sensing lines. Because of this modification, the EGR bypass line installation, and other modifications that were made to the HPCI System during the Outage, the Startup Tests were repeated for this system when the plant restarted in August of 1986. Dynamic Stability checks were again completed using 500 gpa step changes introduced in both manual and automatic flow control modes with the HPCI System operating in a closed loop to the CST. Level 2 criteria was exceeded when HPCI System flow had a measured decay ratio of 0.28 resulting from a mid-flow speed decrease step change in the manual Supplet.snt 9 Pag) 3 13-3 mode. This is currently considered to be acceptable but will be exasined closely in HPCI testing at higher test conditions. During a HPCI automatic initiation in the CST closed loop lineup, a HPCI System flow of 5000 sps was achieved in 21.2 seconds. Twenty-five seconds af ter the automatic initiation occurred, HPCI flow was 5003 sps at 1185 psig pump discharge pressure, 225 l psig greater than the 960 psig rear'.or pressure. j t Data was also taken during this test to determine ; I the actual 300% value for the HPCI steam supply line l high flow isolation trip setpoint. However, the i trip setpoints were not adjustei to these settings, but are being left at the current trip setpcints given in Technical Specifications. The current l isolation settings as specified in Technical l Specifications are considered acceptable as they are l conservative yet provide ample margin to prevent i spurious HPCI isolations on system automatic l initiations. All other Level 1 and 2 criteria were met. During ratesting of HPCI in September of 1986, a sluggish response was noted in the HPCI control valve. In an attempt to make the HPCI Systes more responsive, it was decided to replace the EGR in the hydraulic portion of the HPCI control system. As a result, the 1000 psig hot CST injection was repeated to verify proper control system operation. HPCI was successfully quick started and HPCI discharge flow reached the 100-percent-rated valut (5000 gps) in 21.0 seconds. Following the automatic initiation, HPCI flow leveled out at 5100 gpm with a discharge pressure of 1190 psig. The initial speed peak was 2134 rpm and the maximum peak was 4114 rpm. All other Level 1 and Level 2 criteria were met. l In June of 1987, following the February 1987 turbine rotor replacement (reference LER 87-006-00) and l prior to the scheduled Test Condition Three HPCI l test sequence, tuning of the HPCI governor control system was performed. During this tuning, a RCIC turbine trip occurred on low suction pressure when the HPCI turbine was Quick Started. To prevent recurrence, HPCI and ACIC suctions were aligned to different sources. Supplen:nt 9 P;ge 3 13-4 During the initial vessel injection attempt, the NPCI turbine underwent a total of five overspeed trip /rese;. actions, violating Level 1 critaria, prior to being secured. Two diagnostic CST to CST runs determined the overspeed c0nditions were minimua flow related, and consequently, the second vessel injection attempt was to provide an immediata flowpath to the vessel by manually opening the injection valve immediately following the Quick Start. The second vessel injection attempt was aborted when a logic probles caused the injection valve to cycle closed, creating a water hammer damaging the suction relief valve, suction pressure instrumentation and the flow transmitter. In addition, the RGSC was found to be defective. Following repairs to the suction relief valve and replacement /recalibration of the RGSC, suction and flow instrumentation, retuning was performed. Once the governor control systen had been retuned, a third vessel injection attempt and dynamic stability checks were performed, this time successfully. Time to rated flow was 25.2 seconds, exceeding the Level 1 criteria of 25 seconds. The initial speed peak was 1096 rpm and the maximum speed peak was at 3991 rpm. All speed and flow step changes exhibited acceptable decay ratios. At no time did the gland seal condenser system allow steam leakage to atmosphere. Following the required 72 hour cooldown period, a cold vessel inje: tion attempt resulted in two overspeed trip / reset actions, a Level 1 criteria violation. Per GE recommendation, the control valve hydraulic assist valve was fully closed and retuning was performed. After the retuning effort, another HPCI vessel injection and dynamic stability checks were l performed, resulting in a time to rated flow of 22 3 l seconds with initial and maximum speek peaks of 1222 and 4303 rps, respectively. This exceeded the Level 2 criteria for a maximum speed peak of 4200 rpm. Several speed and flow steps at aid flow conditions failed to achieve Level 2 quarter damping criteria. At no time did the gland seal condenser systes allow steam leakage to atmosphere. Supplement 9 P ge 3 13-5 After the required 72 hour cooldowa period, HPCI was Cold Quick Started to the vessel. Time to rated flow was 27 5 seconds, exceeding the Level 1 criteria of 25 seconds. The initial and maximum speed peaks were 1095 and 4461 rps, respectively. This exceeeded the Level 2 criteria of a maximum speed peak of 4200 rpm. At no time did the gland seel condenser system allow steam leakage to ! atmosphere. l l The second Cold Quick Start to the vessel occurred i 286 hours after the previous Cold Quick Start, far l in excess of the required 72 hour cooldown p?riod. l Time to rated flow was 30.85 seconds, exceeding the ! Technical Specification allowable value of 30 I seconds and the Level 1 criteria of 25 seconds. The l initfal and maximum speed peaks were 2918 and 4328 ; rps, respectively, exceeding Level 2 criteria for a l maximum speed peak of 4200 rps. At no time did the ' gland seal condenser system allow steam leakage to atmosphere. During a diagnostic test to investigate HPCI performance after a 24 hour cooldown period, the HPCI turbine tripped on overspeed. In order to further investigate HPCI performance, five diagnostic HPCI CST to CST test runs wero performed. l As a result of this and other investigations, tra HPCI turbine control oil system was disassembled, cleaned, and inspected and the HPCI EGR was replaced. During the HPCI outage, the HPCI discharge check valve was changed from a lift chec!< i 1 to a swing check in an attempt to improve closing l Limes to mitigate suction piping overpressure I transients observed during HFCI turbine trips. I Folicwing HPCI operability checks, tuning was again performed resulting in acceptable turb'Je j performance. HPCI ...4 ' art performance was ) further improved , chai.2L ; out the HPCI stop valve limit switches, r..%...., .. , delay to the RGSC ramp start. In October of 1987, the Test Condition Three HPCI Vessel Injectica test sequence was reperformed in its entirety, beginning with the Hot Ve>sel Injection. Following a manual start to the vessel, dynamic stability checks were performed. Two of the average flow steps, t 500 gpm at 2200-2700 gpm, did not meet Supplescnt 9 Page 3 13-6 the Level 2 criteria for quarter damping. This condition was accepted because of the high degree of stability at higher flow rates. Following the nanual start a Het Quick Start to the vessel was performed, with rated flow occurring after 20 5 seconds. The maximum transient speed peak was 4117 rpm, All other Level 1 and Level 2 critaria were met. Following a 91 hour cooldown, the first HPCI Cold Quick Start was performed, with rated flow occurring y after 21.5 seconds. The maximum transiint speed peak was 4130 rpm. All other Level 1 and Level 2 criteria were met. The final EPCI Cold Quick Start was performed following a 74 hour cooldown period. The maximum transient epeed peak was 4123 rpr and rated flow was obtained 21.4 seconds after initiation. Approximately one mint : e '.nto the test, the HPCI turbine tripped on High RPV Water Level (level 6). Because of the short duration of the test, Glard Seal Svstem data could J.. t be takea. This Level E criteria violation was accepted based on acceptable Gland Seal System performance on all prior tests. The HPCI turbine trip on Level 8 was avoidable with a more rapid feedwater turbine speed adjustment and was not the result of any HPCI System component malfunction and, therefore, was not considered to be a violation of the Level 1 criteria. All other Level 1 and Level 2 criteria were satist'ied. Fellowing the completion of HPCI Vessel Injection testing, the final Cold CST Quick flart test was , , performed to collect baseline data for the Operations Surveillance Testing Program. After a 72 hour cooldcan period. HPCI was Quick Started to the CST, with rated flow occurring after 19.9 ^*cands. The maximum transien'. speed peak was 414f 4 and stead) state flow stabilizco at 5400 gpm t. 4 discharge pressure at 1060 psig. Data gathered during the above testing to determine the HPCI steam line high flow isolation trip setpcint has been further evaluated by Nuclear , 'eineering. The results of this evaluation date the flow equation used to determine the ;,.ial trip setpoint as pres 3ntly listed in Tech. 4' O. Table 3 3 2 ' and therefore no adjustment is 4 Supple ent 9 Pago 3 13-7 necessary. This evaluation is detailed in Design Calculation #4572 Revision C. ) During the startup following the Spring 1988 LLRT Outage, the final Cold and Hot CST Injections and stability checkJ were performed at a reactor pressure of 150 psig. During the Cold Quick Start, HPCI discharge flow reached Freater than 5000 gpm within 19.2 seconds. Pump discharge pressure was 275 psig which was greater than 100 psig above reactor pressure. HPCI speed reacned a max:aum nalue of 1816 ria which is well below the 10% margin to the overspeed trip setpoint of 5000 rpa (4420 rpm). All applicable test criteria were met. Follo,ing the above testing, HPCI was started manually and flow steps ir, both manual (speed control) and auto (flow control) modes were performed with pump flow between 4500 and 5100 gpm. Stability eas successfully demonstrated by this testirg and all HPCI system variable respenses were shown to have decay ratics less than 0.25. A Hot Quick Start was then performed and HPCI discharge flow reached greater than 5000 gpa within 16.1 seconds with a discharge pressure of 290 psig which was greater than 100 psi above reactor pressure. The maximum speed reached by HPCI was 2816 rpm. All applicable test critetia were met-This concludes all required HPCI testing during the Startup Test Program. Supplement 9 Pcge 3 14-1 3 14 Selected Process Temperatures 3 14.1 Purpose T6.e purposes of this procedure are to establish the proper setting of the low speed limiter for the recirculation pumps to avoid coolant temperature stratification in the reactor pretsure vessel botton nead region, to provide assurance that the seasured bottom head drain temperature corcesponds to bottom head coolant temperature during normal operations, and to identify any reactor operating modes that cause temperature rtratification. 3 14.2 criteria Level 1 The reactor recirculation pumps shall not be restarted nor flow increased unless the coolant temperatures between the steam done and bottom head drain are within 145 F. The recirculation pump in an idle loop must not be started, active loop flow must not be raiseo, and power must not be increased unless the idle loop suction temperature is '#1 thin 50 F of the active loop suction temperature. If two pumps are idle, the loop suction temperature must be within 50 F of the Oteam done temperature before pump startup. Level 2 During operaticn of two recirculation pumps tat rated core flow, the bottom head temperature as mer.sured by tha botton drain line thermocouple should be within 30 F of the recirculation loop temperatures. I 3 14 3 Results For the initial testing conducted in 1985, the l c7olant temperatures measured at 30% Recirculation n.up speed satisfied t'.ie Level 1 criteria. The instability of the recire. speed controller that occurred during this test precluded an effective investigation of the stratificatior, phenomenon at low flows. The test also allowed setting of the low speed limiter based on flow controller variations ofi 2% of rated speed. Flow controller vaciations of i 5% were experienced prior to stratification so the test was terminated. Supplement 9 P ge 3 14-2 The sintaus recirculation pump speed data collection was resumed in August, 1986 following completion of the preceding Outage. In subsequent heatup testing, the Leciro MG Sets were hand cranked down to speeds of about 20%. The Level I criteria was satisfied at all times during this test. The low speed limiter setting was chosen to be 285 speed based on the previously observed controller instability below that level. During Test Condition Six, with the reactor operating at 95.9% CTP and 975 CF, the bottos head temperature al: measured from the botton drain line 0 thermocouple was within 17 F of the rea,irculation loop temperature thereby satisfying the Level 2 criteria of 5 30 F delta temperature. Following entry into and completion of Test Condition 4 testing, tte differential temperature between the steau done and the botton heed temperature was verified to be within 440 F prior tc the restart of the first recirculation pump, thereby satisfying the Level 1 criteria that this difference be < 145 F. Additionally, the loop suction temperature was verified to be within 1 F0 of the steam done temperature prior to this first pump restart which satisfies the Level 1 criteria that this difference be 5 50 F. Although the second pump was not restarted due to the failure of its disc %rge valve to close and permit restart, the suction temperature of this idle ; loop wcs verifit.d to be within 3 F of the active l < loop suction temperature and therefore would have satisfied the Level I criteria that this difference ' be 5 50 F had the pump been restarted. This concludes the required selected process , temperatures section of the Startup Test l'rogram. . 4 Supplement 9 Pag 3 3 15-1 3 15 Systen ExpaAston 3 15.1 Purpose The purpose of this test is to verify that selected plant piping systems are free and unrestrained with regard to thermal expansion, and to verify that the thernal movement of the piping and associated support system components is consistent with the analytical prediction of the piping system stress analysis. 3 15.2 criteria Level 1 The seasured displacene'1ts at the instrument *d locations shall be within the greater of the specified allowable tolerance of the calculated values, or i O.25 inches for the specific points. There shall be no obstruction thich will interfere with the expected thermal expansion of the piping system. Electrical cables shall be able to accommodato expected thermal expansion of the piping system. Instrumentation ano' branch piping can accommodate expected thermal expanelon of the piping system. The constant hanger shall not be bottomed or tocped out. The spring hanger ahall not be bottomed or topped out. The snubber shall not be bottomed or topped out. Level 2 The measured displacements at the instrumented locations should be within the greater of the specified expected tolerance of the calculated values, or 1 0.25 inches for the specific points. The installed cold posit'on of the constant hanger must be within 1 55 of the design cold load. The installed cold position of the spring hanger must be within 1 5% of the design cola load. Sapplement 9 ' P:g3 3 15-2 The snubber may doviate from its design cold position setting + 1/2", providing the position is not less than 1/2n from bottoaing out. 3 15 3 mesults Piping Inspection Results Selecte.' piping systems were walked down at various plant conditions to identify possible restraints to projected thermal expansion. These walkdowns occured at ambient temperature, 2500F and rated temperature. Hanger and snubber settings were recorded ar.d thermal expansion (PVDET) sensors were verifled to be intact. No restraints to projected thermal expansion were identified. One-hundred and forty-three (1883) supports were identified as being out of tolerar.ce or topped or bottoned out. Following re-verification and engineering evaluation, sixteen (16) supports were adjdsted or modified and the remainder accepted as is. The East and West Main Steam Bypass Lines were replaced during the Outage which started in the fall of 1985, because of cracks which were discovered in these lines. During subsequent testing following reactor restart in August, 1986 these lines were visually inspected to verify that they were unrestrained with regards to projected thermal r expansion. These walkdowns occured at ambient temprature;andatrecirclooptemperaturesof 350 and rated. No restraints to bypass line thermal expansion were identified. Five supports were found out of tolerance, and upon eneineering evaluation were accepted as-is. Third therst.A cycle visual inspections and hanger readings were made on all system piping including the replaced Main Steam Bypass Lines. There were no restraints to thermal expansion identified. Two-hundred-ninety-five (295) supports were identified as not being within their proper working range. Following engineering evaluation and reverification, eight (8) suppc t vere reset and the remaining supports accJpt- es-is". Supplement 9 Page 3 15-3 System Espansion Results Selected points on the piping systems were wired with r uote sensors to monitor the thermally induced piping movements during systes operation. The monitored points were espected to undergo large movements or experience large thermal stresses. After establishing initial readings for the sensors at sabient. conditions, the sensors were monitored during thr .nitial heatup of the plant. Data was recorded at 507 intervals until the reactor reached operating temperatura. The evaluations
- 4. found several criteria eroeedances, but upon engineering evaluation of the exceedances, all were found acceptable.
In adJitior., initial ambient sensor readings taken before Heatup were compased to ambient sensor readings after a Heatup and cooldown cycle was completed. No appreciable difference in the before and after readings were noted, indicating piping movement was not restrained. Thermal Espansion data was again taken at M F intervals at moderator temperatures beginning at 1000 F during the subsequent heatup cycle following initial heatup. The data was evaluated at each temperature plateau before proceeding to the next level. Upon reaching rated temperature, four Level 2 criteria violations existed, but these were very minor ami accepted as-is. The East and West Main Steam Bypass Lines that were replaced in the fall of 1985 were also monitored for trupected thermal expansion during the subsequent heatup after the Outage. The heatup and cooldown senor readings satisfied all Level 1 and Level 2 criteria except at the 350'F ree bc loop temperature plateau. At that point there was one Level 2 failure whicit resulted from inadequate heating of the bypass piping due to the bypass valves being closed at the time the test was performed. At higher temperatures data was taken witI the bypass valves open, and all criteria were satisfied. During Test Condition Six, with the reactor operating at 96.6% CTP and 97 5% CF, additional sensor readings were taken. Movement was determined from baseline readings taken at cold conditions during the previous plant shutdown. Supplement 9 Page 3 15-4 There were two apparent Level I criteria exceedar.ces and twenty three Level 2 criteria exceedances associated with this data collection. Displacement sensor D-203 located on the RCIC Steam Supply Line had a reading of 580 mils vs an allowable of 241 mils; however, sensors D-201, D202 and D-204, also located on the RCIC Steam Line, were very close to the analytical prediction. Additionally, displacements of sensors on the B main steam line are close to their analytically predicted displacement which indicates that the overall header is moving in the predicted direction end therefore sensor D-203 any not be working properly; however, , if this is net the case, preliminary calculations show that the marisua stress in the RCIC system caused by this exceedance is 7% of the allowable stress and is therefore acceptable. i The load measured by force sensor K004A was 28,321 lbs vs an allowable Level 1 value of 12,515 lbs. ! Previous to this test during initial heatup of the : plant in 1985, a load of 35,687 lbs. was measured by this sensor and detailed calculations at that time deemed that load to be acceptable. Since the present load is less than that previously evaluated and found acceptable, this load is also acceptable. Detailed evaluations by Sargent & Lundy are ongoing for the above Level 1 violations and the twenty three Level 2 violations but not yet complete as of tisis report date. I f Supplement 9 P ge 3 16-1 3 16 Core Power Distribution NOTE: As discussed in memorandus VP-86-0141, "Startup Test Program Changes", dated October 17, 1986, from Frank E. Agosti to James G. Keppler, it is our intention to delete this test. I l f I i i i Supplement 9 Pag 3 3 17-1 3 17 Core Performance 3 17.1 Purpose
- a. To evaluate the core thermal power.
- b. To evaluate the following core performance parameters
- 1. Maximum linear heat generation rate (MLHGR)
- 2. Miniaun criiical power ratio (MCPR) 3 Maximum average planar linear heat generation rate (MAPLHGR).
3 17 2 crii,eria Level 1 l The maximum linear heat generation rate (MLHGR) ; during steady-state conditions shall not exceed tha ! allowable heat flux as specified in the Technical l Specifications. The steady-state minimum critical power ratio (MCPR) shall be maintained greater than, or equal to, the i l value specified in the Technical Specifications. 1 The maximum average planar linear heat generation I rate (MAPLHGR) shall not exceed the limits gl'ien in I the plant Technical Specifications. Steady-state reactor power shall be limited to full rated maximum valuec on or below the design flow control line. Core flow should not exceed its rated value. Level 2 None 3 17 3 neaults BUCLE computer analyris of whole core TIP traces obtained at 15.6% reactor power showed that all criteria were net, during Test Ccndition One. The Core Perforanance parameters during Test Condition Two were determined using the Process Computer programs P1 (Periodi:: Core Evaluation) and I Supplement 9 Pag) 3 17-2 OD-3 (Core Thermal Power /APRM Calibration). All Level 1 criteria were satisfied upon the , determination and verification of the following parameterst Core Thermal Power (CMWT) Percent of Rated Core Thermal Power (FCT PWR) Core Flow (WT) Maximus Linear Heat Generation Rate (MLHGR) Minimus Critical Power Ratio (MCPR) Marinus Average Planar Linear Heat Generation Rate (MAPLHGR) During Test Cundition Threv, the Process Computer programs (P1 and OD-3 Option 2) were again run to detormine the above parameters: The Process Computer edits werc utilized to determine that all requirements associated with the test were satisfied as follows: 'the Core Marinum Fractien of Limiting Power Density was 0.43 which satisfied the seceptance criteria that this value be less than cr equal to 1.0. The Cere Maximus Fraction of the Limiting Critical Power Ratio was 0.44 which satisfies the acceptance criteria that this value be less than or equal to 1.0. The Core Maximum Average Planar Linear Heat Generation Rate Ratio was 0.42 which satisfies the acceptance criteria th,;t this value be less than or equal to 1.0. The rated maximum value far reactor power at 95 3% of rated core flow was determined to be is 96.5% of rated Core Thermal Power based on the design flow control line. The actual calcult,ted CTP was 48.6% which was below the design flow control line. Measured core flow was 95 3% et rated core flow ! which satiefies the criterik, that core flow does not exceed its rated veluc. During Test Condition Five, the Process Computer programs (P1 and OD-3 Option 2) .4e'.e run during the performance of the Reactor Engirsering procedure 54.000.07 (Core Performance Parameter Check). The Process Computer edits were utill:ed to determine Supplement 9 Page 3 17-3 that all requirements associated with the test were satisfied as follows. The Process Coeputer value of Core Maximun Fraction of Limiting Power Density was 0.663, which satisfies the scoaptance criteria that requires this value to be less than or equal to 1.0. The Process Computer value of Core Maximus Fraction of the Limiting Critical Power Ratio was 0.704, which satisfies the acceptance criteria that requires this value to be less than or equal to 1.0. The Process Computer value of Core Maximus Average Planar Linear Heat Generation Rate Ratio was 0.665, which satisfies the acceptance criteria that requires this value to be less than or equal to 1.0. The rated maximum value for reactor power at 61.4% of rated core flow was determined to t,e 74% of rated Core Thermal Power based on the design flow control line. The actual CTP was 71.8% which was below the ; design flow control line. Measured core flow was 61.4% of rated core flow which satisfies the criteria that core flow does not oxceed its rated value. During Test Condition Six, Process Computer Programs i P1 and OD-3 option 2 were run during the performance of the Reactor Engineering procedure 54.000.07 (Core Performance Parameter Check). The Process Computer edits were utilized to determine that all requirements ascociated with the test were satisfied I as follows: The Process Computer value of Core Maximun Fraction of Limiting Power Density was 0.877. 2 1s satisfies the acceptance criteria which requires this value to be less than or equal to 1.0. The Process Computer value of Core Max!aus Fraction i of the Limiting Critical Power Ratio was 0.849. This satisfies the acceptance criteria which requires this value to be less than or equal to 1.0. r The Proco,s Computer "alue of Core Maxinua Average Planar Linear Heat Generation Rate Ratio was 0.861. This satisfies the acceptance criteria which requires this value to be less than or equal to 1.0. Supplement 9 Pago 3 17-4 The rated maximum value for reactor power at 99.8% of rated core flow was determined to be 100% of rated Core Thermal Power bassa on the design flow control line. The actual CTP was 98.4% which was below the design flow control line. Measured core flow was 99.8% of rated core flow which satisfies the criteria that core flow does not exceed its rated value. { I During Test Condition Four, Process Computer l Programs P1 and OD-3 Option 2 were run during the I performance of the Reactor Engineering procedure I l J4.000.07 (Core Performance Parameter Check). The ! Process Computer edits were utilized to determine l that all requirements associated with the test were satisfied as followst The Process Ccaputer value of Core Marlaus Fraction of Limiting Power Density was 0.39. nis satisfies , the acceptance criteria which requires this value to l be less than or equal to 1.0. The Process Computer value of Core Marinus Fraction of the Limiting Critical Power Ratio was 0.582. This satisfies the acceptance criteria which requires this value to be less than or equal to 1.0. The Process Computer value of (bre Maximus Average Planar Linear Heat Generation Rate Ratio was 0 375. This satisfies the acceptance criteria which l requires this value to t:e less than or equal to 1.0. ! The rated aanlaus value for reactor power at 35% of l rated core flow was determined to be 50% of rated Core Thermal Power based on the design flow control line. The actual CTP was 39.6% which was below the l design flow control line. l Measured core flow ws.s 35% of rated core flow which satisfies the criteria that core flow dues not exceed its rated value. I t Supplement 9 PCse 3018-1 ( 3.18 Steam Production This test was previously deleted from the FSAR (Section 14.1.4.8.18). 1 M . I i I l l I e I f t i 4 1 3 i 1 I l r t r r i i I i, 6 Supplement 9 P ge 3 19-1 3 19 Core Power-Void Mode Response , NOTE: As discussed in memorandum VP-86 0141, "Startup Test Program Changes", dated October 17, 1986, from Frank E. Agosti to James G. Keppler, it is our intention to delete this test. l t i ! j i l r + f i l l l 1 - 1 i i t h J Supplement 9 9 Pag) 3 20-1 3 20 Pressure Regulator 3 20.1 Purpose The purpose of this test is to:
- t. Determine the optimum settings for the pressure control loop by analysis of the transients induced in the reactor pressure control system by means of the pressure regulators.
- b. To demonstrate the takeover capability of the backup pressure regulator on failure of the controlling pressure regulator and to set spacing between the setpoints at an appropriate valut,
- e. To demonstrate smooth pressure control ,
transition between the control valves and bypass valves when the reactor generates more steam than is used by the turbine. 3 20.2 Criteria Level _1_ The decay ratio aust be less than 1.0 for each process variable that exhibits oscillatory response to pressure regulator changes. Level 2 In all tests the decay ratio must be less than or equal to 0.25 for each process variable that exhibits oscillatory response to pressure regulator changes when the plant is operating above the lower limit setting of the master flow controller. I Pressure control deadband, delay, etc., shall be [ small encash for steady-state limit cycles, if any, ; to produce turbine steam flow variations no larger ! than 0.5 percent of rated flow. l During the simulated failure of the controlling pressure regulator along the 100 percent rod line, the backup regulator shall control the transient so [ that the peak neutron flux ne peak vessel pressure remainsbglowthescraasettingsby7.5percentand 10 lb/in. , respectively. i l [ i i , . . , . .--__- - - .- _ - = . - - .-_- - - - _ - _ , . I Supplement 9 : Page 3 20-2 [ After a pressure setpoint adjustment, the time between the setpoint change and the occurrence of the pressure peak shall be 10 seconds or less. , (This applies to pressure setpoint changes made with the recirculation systes in the master or local ! manual control mode.) f 3 20 3 Results j Proper pressure regulator operation was demonstrated j in Test Condition One by analysis of systes response j to step increases and decreases in pressure demand , with the bypass valves open and generator not on the i line. Ad .tional steady-state measurements were i taken with the generator loaded and bypass valves ; closed. All Level 1 and Level 2 criteria were net. I i De pressure setpoint changes on each regulator, r while significant in magnitude (11-13 psig), were ! stable and well damped. At. such no systen tuning was performed in this test condition. [ 3e Regulator failure tests yielded significantly ! different responses (14 pais change for failure of ! #1; 6 pais change for failure of #2). This l discrepancy in response is likely attributable to differences in the time delay circuitry for each channel in the High Value Gate and difference of 1.7 . psig in the sensed pressure being fed to each ( regulator channel. The time delay component in the t regulator high value dates has since been removed. l The testing performed for the Pressure Regulator j during Test Condition Two consisted of introducing ! 10 psig step change and staulated regulator failures l
- j. In the Pressure Control Systes. j
' l The Level 1 criteria for this test during Test l l Condicion Two was satisfied when no process
- variables were found to be divergent and all decay l rntios were less than 1.0 during the 10 psig step ,
changes and uinulated regulatoe failures, j i Steady-state steam flow variations were monitored by i neasuring generator electrical output limit cycling due to pressure controller operation. The Level 2 criteria requiring that these variations are no l larger than 1.0 percent peak-to-peak of rated flow ; was satisfied by analysis of the generator output , which showed a maxinua variation of 0 9 percent ; peak-to-peak c
- rated flow. ;
l 1 ) Supplement 9 '8 age 3 20-3 : 4-The other Level 2 criteria associated with this test 1 required that, after a pressure setpoint adjustest t, j the time between the chenge and the occurrence Of ; the pressure peak shall be 10 seconds or less. l Analysis of this test's 10 pois steps showed peak ; pressures betweer 3 6 and 5 2 seconds, satisfying ; 1._ the criteria. ! I Finally, the elimination of the time delay to backup i regulator takeover resulted in significant laprovement over Test CoM ition One results in response to both normal transfers t'M regulator failures. At no time did the bypass valves enter their "FAST" mode and all transients were controlled and strongly damped. e Pressure regulator testing during Test Condition Three was performed at 71.4 CTP to verify the optimum settings for the pressure control loop by analysis of systes response to step decreases and increases in ptassure demand. In addition, the takeover capability of the backup pressure regulator upon failure of the controlling pressure regulator i was demonstrated. Proper pressure regulator ! operation was demonstrated with both the Turbine 4 Control Valves a, lone and with "incipient" l conditions, defined as that condition where the load 1 demand has just barely closed the bypass valves. - Additional steady-state seasurements were taken with ! the generator loaded and the bypass valves closed. S ' For the 10.0 psig down and up steps and regulator , failures performed in this test, no process J variables were found to be divergent and all decay j ratios were less than 1.0, thereby satisfying the Level 1 criteria. An analysis of the steady state generator output data recorded during this test shows a maximum peak to peak value of 9 3 We which is less than 11.54 We (15 peak-to-peak) and, therefore satisfies the Level 1 criteria. An analysis of the 10.0 psig down and up steps show peak pressures occur between 4.3 and 6.0 seconds after strip initiation, well within the ten (10) second Level 9 criteria for this test. Supplement 9 Pag 3 3 20-4 'Ihe followi';.6 Table suamarizes the pressure data resulting from the step changes and regulator failures Initial Max / Min Final Test Description Press Press Press Icy Reg #1 Failure 963 7 973 5 967.2 Reg 42 Failure 963 7 973 7 967.5 Reg #1 10 psig downstep 963 5 951.0 952.9 M g #1 10 psig upstep 952.9 965.6 963 7 Rep #2 10 psig dot ' step 964 3 951.7 953 1 Reg #2 10 psig upatep 953 3 966.0 963 9 Incipient Reg #1 10 psig downstep 963 9 953 3 953 9 Reg #1 10 peig upstep 953 7 964.8 963 5 Reg #2 10 psig downstep 964 3 953 3 954.2 Reg #2 10 psig upstep 954.2 965.4 964 3 Pressure regulator testing during Test Condition Five was performed to verify the optinua settings for the pressure control loop by analysis of systet response to step decreases and increases in pressure demand. Proper prassure regulater operation was demonstrated with both the Turbine Control Valves and the Bypass Valves controlling pressure. Additional steady-state measurements were taken with the generator loaded and bypass valves closed. For the to psis down and up steps performed in this test, all process variables were strongly damped and no decay ratios were found to exceed 0.25, satisfying both the Level 1 and Level 2 criteria. An analysis of the 10 psig down and up steps show : peak pressures occur betwer' 3 2 and 5.1 seconds ; after step initiation, well within the ten (10) second Level 2 criteria for this test. L Rated turbine steam flow is equivalent to 1154 MWe, ! consaquently, il peak to peak variations aust be less than 11.54 NWe. An analysis of the steady I state generator output shows a marinun peak to peak f value of 9 928 MWe, thus satisfying the Level 2 criteria. , Supplement 9 Page 3 20-5 The following Table summarizer the pressure data resulting from the pressure setpcint step changes. Initial Max /Hin Final Test Description Press Press Press TGV Stepu Reg #1 10 psig downstep 964.8 952 3 953.9 Reg #1 10 psig upstep 953 7 966.8 964.7 Reg #2 10 psig downstep 964.5 952.1 953 4 Reg #2 10 psig upstep 953 7 966.8 964.5 LPV Steps Reg #1 10 psig downstep 965.0 953 3 954.5 Reg #1 10 psig upstep 955.8 968.3 965 2 Reg #2 10 psig downstep 965.2 953 1 954.4 Reg #2 10 psig upstep 954.5 967 7 965.6 Pressure Regulator testing has not yet been completed in Test Condition Six, i Supplement 9 . Page 3 21-1 3 21 Feedwater System 3 21.1 Purpose
- a. To adjust the feedwater control system for acceptable reactor water level control.
- b. To demonstrate stable reactor response to subcooling changes.
- c. To demonstrate the capability of the automatic core flow runback feature to prevent low water level scram following the trip of one feedwater pump.
- d. To demonstrate adequate response to feedwater heating loss.
- e. To determine the maximum feedwater runout capability.
l 3.21.2 criteria Level 1 The response of any level-related variable to any test input change, or disturbance, must not diverge during the setpoint changes. For the feedwater temperature loss test, the maximum feedwater temperature decrease due to a sir.qle failure case must be less than or equal to 100 0F. The resultant MCPR must by greater than the fuel thermal safety limit. For the feedwater temperature lass test, the increase in simulated heat flux cannot exceed the predicted Level 2 value by more than 2 percent. The predicted value will be based on the actual test values of feedwater temperature change and power level. The feedwater ficw runout capability must not exceed the assumed value in the FSAR. Level 2 Level control system-related variables may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0.25, as a reselt of the setpoint change testing. Supplement 9 Pag) 3 21-2 A scram must not occur from low water level following a trip of one of the operating feedws.ter pumps. There should be a greater than 3-in. water-level margin to scram for the feedwater pump trip. For the feedwater temperature loss teht, the increase in slaulated heat flux cannot exceed the predicted value referenced to the actual feedwater temperature change and power level, which will be taken from the Transient Safety Analysis Design Report. The average rate of response of the feedwater actuator to large (>20 percent of pump flow) step disturbances shall be between 10 to 25 percent of pump rated feedwater flow /sec. This average response rate will be assessed by determining the time required to pass linearly through the 10 percent and 90 percent response points of the flow transient. The dynamic flow response of each feedwater actuator (turbine or valve) to small (<10 percent) step disturbances shall be the followingt
- a. Maximum time to 10 percent of a step disturbance
$1.1 sec.
- b. Maxinua time from 10 to 90 percent of a step disturbance $1 9 sec.
1
- c. Peak overshoot (percentage of step disturbance) 115 percent.
3 21 3 nesuits During the initie.1 heatup, the feedwater system l performed satisfactorily in both the manual and automatic modes. All level-related variables did not diverge during testing and all system related variables did not exceed a 0.25 decay ratio for their oscillatory responses in the level setpoint changes. All applicable test criteria were satisfied. During Test Condition One, as previously done during the heatup testing, the Startup Level Controller setpoint was adjusted to simulate step changes of three inches for Reactor water level. During the setpoint increase water level increased in a smooth > supplement 9 Page 3 21-3 ^ manner with little overshoot and stabilized within 75 seconds. During the setpoint decrease water : level decreased and overshot the three inch down step by 2 to 3 additional inches. This overshoot , dampened rapidly and water level stabilised within
- 110 seconds.
The Test Condition One test was completed ; satisfactorily. The criteria that the decay ratio ! of level control systea-related variables being less j than 25 was met for all portions of this test. During Test Condition Two, feedwater system testing . was limited to single element master level i controller step changes due to equipment problems I with the Dynamic Compensator Lead / Lag Network
- Computation Module. The dynamic flow response of i the Reactor feed pump turbines was not able to be !
i checked because the flow to the Reactor was i insufficient to allow aut.oaatic 16 vel control with ; two pumps operating witt both minimus flow bypass [
- valves shut. Both minimum flow bypass valves are ;
required to be closed to adequately measure the flow , response of the feedwater actuators to step inputs. l Feedwater systes response to five inch Reactor level ' changes using setpoint tape manipulations in single ! element automatic control were smooth and l j controlled. All applicable acceptance criteria were . 2 set for the conditions tested. l l In Test Condition Three, at a reactor power of 485, ! testing was conducted in both One Element and Three ! Element *4 odes, with each feedpump feeding the vessel I and the o:her in standby. This satisfied the above f noted Test Condition Two testing that could not be : i completed earlier due to the inoperative Dynamic , ; Computation module, j Both SarPT Control Systems (Systea #1 and System #2) l were tuned and + 105 speed demand steps with the l pump in the recirculation mode were performed. j i After the completion of SRFPT Speed Control System 4 testing, the NRFP was then placed in standby after l ! the SRFP was placed into service feeding the j vessel. Level setpoint tape changes of up to + 5 , i inches were performed in both One E w ent and Three ! Element modes. Once proper Level ContN1 System j . response was verified, the 1 5 inch level setpoint ; I adjustment ramps were performed in both One and ; ) Three Element modes. l i i f 4 l l } ! t Supplement 9 Page 3 21-4 Following completion of SRFP testing, both of the NRFPT Speed Control Systems were tuned and tegted, again with the pump in the recirculation mode. Once the KRFP was placed in service feeding the vessel, level setpoint change testing was performed in the same manner as the SRFP. The STARTREC traces for both One and Three Element Control mode were analyzed for quarter damped response. The following signals were deemed to be
- Level Control System-related:
Feedwster Control Function Generator Output - NRFP Feedwater Control Function Generator Output - SRFP Master Feedwater Controller Output North Reactor Feed Pump Flow South Reactor Feed Pump Flow North RFPT Speed South RFPT Speed All of the above signals showed quarter damped (0.25) response to i 5 inch level setpoint changes which satisfies the Level 1 criteria of non-divergence and the Level 2 criteria of decay ratio. In Test Condition Three at a reactor power of 71.4%, additional water level setpoint changes (1 5 inches) in both Single and Three Element modes were performed. The applicable Level 1 criteria for no divergence and Level 2 criteria for quarter damping i were set for the testing performed. Planned testinc to verify the dynamic flow response and rate of response and the feedwater turbine actuators (Level 2 criteria) could not be performed due to Feed Pump Turbine speed control and hydraulic control oil system problem. In Test Condition Five at a reactor power of 71.25, water level setpoint changes (t 5 inches) were again performed in both Single and Three Element modes. The Level 1 criteria for no divergence and the Level 2 criteria for quarter damping of level control systes related signals were met. This testing provided confidence that the Feed Pumps would adequately respond to expectea desands until ! hydraulic control oil systen repairs and turbine j speed control systen modifications could be made during the Spring 1988 LLRT Outage, l i Supplement 9 Page 3 21-5 Prior to the shutdown for the Spring 1988 LLRT Outage, diagnostic tests were performed to determine the physical response of the RFPTs to open loop step changes. This response information resulted in a , redesign of the Woodward Governor control amplifier cards, one of which was installed in the NRFP Turbine Systes #2 on an experimental basis. During the LLRT outage, th- Governor pilot actuators on both RFPTs were replaced with new units and the control oil systems were modified to install . hydraulic accumulators. The entire feedwater/governce control system was also recalibrated on both RFPTs. , Due!ng the power ascension in the startup following the LLRT outage, limited inner speed loop (1 60 rps) step response testing of the original and esperimental speed control amplifiers verified that the newly modified amplifier cards would be necessary since speed control stsbility could not be achieved with the original amplifier cards. An experimental amplifier card was then also installed in the SRFP Turbine Systen #2 and stability of the feedwater control system was demonstrated by the performance or + 60 rps inner speed loop steps on ~ both RFPTs and 1 5 inch level controller setpoint changes in 3 element mode only. Following the completion of Test Condition Six steady state testing during the startup after Outage 88-02, the entire Test Conditien Three feedwater tuneup optialzation and test sequence were performed
- at approxiantely 75% CTP with the newly designed l speed control amplifier cards installed in both J speed control channels of both the North and South i RFPTs.
4 i t Supplen:nt 9 Page 3 21-6 The results of the speed control system testing is tabulated below. Dynamic Flow Response to small (< 10%) step I disturbances (100 rps = - 7% flow) ! l Delay Rise Peak Settle Step Tia* Time Overshoot Time Systen Size (Sec) (Sec) (%) (Sec) RFPT Number (rpa) $ 1.1 3 1.9 5 15% 5 14 North #1 100 up 0 36 0.80 23 4.14 North #1 100 dn 0.52 0.70 23 5.96 North #2 100 up 0.44 0 90 18 4.42 North #2 100 dn 0.48 0.84 22 5.06 South #1 100 up 0.46 0.80 23 5 32 South #1 100 dn 0.40 0.74 20 4.66 South #2 100 up 0 50 0.80 23 3 30 South #2 100 dn 0.52 0 76 21 5.76 As can be seen from the above, all applicable Level 2 criteria with the exception of Peak Overshoot were met; however, this has been deemed acceptable by OE based upon the acceptable rate of response to the steps and the high degree of stability following the steps. Average Rate of Response to large (> 20% flow) disturbance 250 rom = 215 flow) Rate of Response Systen Step Size $ Flow /Sec RFPT Number (rpa) >10 <25% Flow /Sec North #1 250 up 20.0 North #1 250 dn 15.6 ' North #2 250 up 18.0 North #2 250 dn 17.2 South #1 250 up 18.8 South #1 250 dn 12.1 South #2 250 up 18.0 South 82 250 dn 11 7 As can be seen from the above, the Level 2 criteria has been met. I i i . - . , . - . . - - . , - . . - - _ _ _ _ - - - _ , . - . . _ - -._ _ ,__ _ - - -__ . -,_ ,~, , Supplement 9 Pago 3 21-7 Following the completion of the above RFPT tuning, t 5 inch Reactor Water level setpoint changes were performed with the feedwater controller in both 3 element and single element modes. The Level I criteria for non divergence and the Level 2 criteria for quarter dampening of all level control related variables was set. With the reactor operating in Test Condition Six at 96.8% CTP and 97% CF, ! 5 inch Reactor Water level setpoint changes were again performed with the feedwater controller in both 3 element and single element modes. The applicable Level 1 and Level 2 criteria were net. The remaining feedwater system testing required for the Startup Test Program which includes the Maximus Feedwater Runout Capability, One Pump Trip and Loss of Feedwater Heating tests has not yet been i performed. t i k 4 i I i l - - - . . . - - , - - -n, , ,,, .,. ,,--.,,- ,----n . . , , - . - - - - , ,- , - - - - - - . - - - - - , , . , - . . , . - , , - - , ~ - . . . _ - _ , , - ,- - , Supplement 9 Page 3 22-1 3 22 Turbine Valve Surveillance - 3 22.1 Purpose To demonstrate acceptable procedures and maximum power levels for surveillance testing of the main turbine control and stop valves without producing a reactor scram. 3 22.2 criteria Level 1 None Level 2 Peak neutron flux aust be at least 7 5 percent below the scram trip setting. remainatleast10lb/in.geakvesselpressuremust below the high-pressure scras setting. Peak heat flux aust remain at least 5.0 percent below its scram trip point. Peak steam flow in the high-flow lines must remain 10 percent below the high-flow isolation trip , settings. 3 22 3 nesults Turbine Control and Stop Valve Surveillance testing has been performed up to a power level of 91.4%
- CTP. All criteria to that point have been satisfied and from a reactor physics standpoint this test could be performed at higher power levels; however, due to a balance of plant considerations with fluctuating heater levels and turbine control valves nearing 100% open, the highest power level reconsended to perform this test was determined to be $ 90% CTP. ,
i Suppler.nt 9 P ge 3 22-2 The results of the testing performed are tabulated below for the most limiting control valve. Margin to Margin to Margin to Margin to Neutron High Heat High Rx Flux Pressure Flux Steam Power Scram Scram Scram Flow (%)_ (t) (psi) (%) (Mlb/hr) Acceptance 375 Criteria 3 10 250 3 0 354 STUT.050.024 70.8 40.2 95.8 8.68 2 34
- STUT.050.024 Supplement 1 '76.8 33 8 93 9 13 29 2.03 <
Information Only Data Set 80.0 33 0 91.0 8.00 1.98 STUT.050.024 Supplement 2 84.5 29.2 84.8 7 34 1.40 STUT.050.024 Supplement 3 87.8 21.6 80.0 6.98 1.17 STUT.050.024 Supplement 4 90.0 22 3 80.0 9 52 1.11 : STUT.050.024 Supplement 5 91.4 21.8 72.7 9 14 1.08
- This data set taken was not on the 100% rod line, resulting in the higher margin to heat flux scram (flow biased).
Additionally, the East and West Bypass Valves were , tested at 70.8% CTP; however, the criteria, although met at this power level, does not apply to this testing since the valves are manually stroked slowly open then slowly closed in turn and at no time do , the valves enter a fast open/ trip closed mode. Therefore, there is no transient associated with pressure regulator response, and the only effect seen is due to the reduction in feedwater temperature due to the diversion of - 155 steam flow to the condenser. A confirmatory Bypass Valve test , at approximately 905 CTP is planned to evaluate the ! effects of Bypass Valve testing on balance of plant ' equipment at that power level. 1 Supplca:nt 9 Pag) 3 23-1 3 23 Main Steam Isolation Valves 3231 Purpose
- a. To check functionally the main steam line isolation valves (MSIVs) for proper operation at selected power levels,
- b. To determine reactor transient behavior during and after simultaneous full closure of all MSIVs.
- c. To determine isolation valve closure time.
3232 criteria Levelj The KS5V stroke time (ts) shall be no faster than 3 0 seconds (average of the fastest valve in each steamline) and for any individual valve 2.5 seconds its <5 seconds. Total effective closure time for any Individual MSIV shall be tsol plus the maximum instrumentation delay time and shall be 15.5 seconds. The positive change in vessel done pressure occurring within 30 seconds after the simultaneous full closure of all MSIVs must not exceed the Level 2 criteria by more than 25 psi. The positive change in simulated heat flux shall not exceed the Level 2 criteria by more than 2 percent of rated value. Flooding of the nain steam lines shall not occur following the full MSIV closure test. The reactor must scram during the full simultaneous MSIV closure test to limit the severity of the neutron flux and simulated fuel surface heat flux transient, Level 2 During full closure of individual valves, peak vessel pressure must be at least 10 psi below scram, peak neutron flux must be at least 7.5 percent below scram, and steam flow in individual lines must be at least 10 percent below isolation trip setting. The peak heat flux must be at least 5 percent less than its trip point. The reactor shall not scram or isolate as a result of individual valve testing. l Supplement 9 Page 3 23-2 ' The relief valves must reclose properly (without leakage) following the pressure transient resulting from the simultaneous MSIV full closure. The positive change in vessel done pressure and simulated heat flux occurring within the first 30 seconds after the closure of all MSIV valves must not exceed the predicted values in the Transient Safety Analysis Design Report. Predicted values will be referenced to actual test conditions of initial power level and done pressure and will use beginning of life nuclear data. The predicted values will be corrected for the appropriate sensured parameters. After the full MSIV closure, the initial action of the RCIC and HPCI shall be automatic if L2 is reached, with RCIC capable of establishing an : average pump discharge flow equal to or greater than 600 gpa within the first 50 seconds after automatic initiation and HPCI capable of establishing an average pump discharge flow equal to or greater than , 5000 spa within the first 25 seconds after automatio initiation. If the low-low set pressure relief logic functions after the simultaneous full MSIV closure test, the l open/close actions of the SRVs shall occur within 120 psi of the low-low set design setpoints. The total number of opening cycles, for the safety / relief valves opening on low-low setpoint, after initial blowdown is not to exceed four times during the initial 5 minutes following isolation. If any safety relier valves open as a result of this test, only one valve may re9 pen after the first blowdown. Recirculation pump trip shall be initisted if L2 is reached after the MSIV full closure test. 3233 nesults During the Heatup Test Condition, with the RPV at rated terperature and pressure conditions, each of the inboard and outboard isolation valves were successfully closed slowly to the approximately 90% open position and then fully reopened, without any noticeable change in reactor pressure, APRM readings or reactor water level. l Supplement 9 P ge 3 23-3 In Test Condition One, with the Reactor at 75 power, a fast full closure af each individual MSIV was performed. All applicable Level 1 and Level 2 criteria were met. The closure times are shown in the table below, using a calculated maximum instrument delay time of 0.299 seconds. Test Condition One' l I I I I l H51V l ts I tsol l Total l l l 1 1 I l F022A I 4.298 1 4.611 1 4. 910.,_ ] l F022B l 3.505 1 3 703 1 4.002 1 , 1 F022C l 4.798 l 4.904 l 5.203 l l F022D I 3 205 I 3 301 1 3.600 l l F028A l 4.294 I 4. 387 l 4.686 1 l F028B l 3.809 I 3.839 I 4.138 l l F028C l 3.617 1 3.899 I 4.198 l , 1 F028D I 4.057 1 4.226 1 4.525 l i # All recorded times are seasured in seconds. During Test Condition Three, with the reactor at 69.2% CTP, each of the inboard and outboard isolation valves were successfully closed slowly to the approximately 90% open position and then fully . reopened without any noticeable change in reactor i pressure, APRM readings or reactor water level. ; i ) During Test Condition Six with the reactor at 96.8% l CTP, each of the inboard and outboard isolation valves were successfully closed slowly to the approximately 90% open position and then
- Ally reopened without any noticeable change in reactor pressure, APRM readings or reactor water level.
i Thr remaining Level 1 and Level 2 criteria are associated with the MSIV simultaneous full closure and will be verified when that test is performed. 1 I Supplement 9 Pag] 3 24-1 3 24 Relief Valves 3 24.1 Purpose The purposes of this test are to verify that the Safety Relief Valves (SRV) function properly (can be opened and closed manually), reset properly after operation, and that there are no major blockages in the relief valve discharge piping. 3 24.2 Criteria Level 1 There should be a positive indication of steam discharge during the manual actuation of each valve. Level 2 Variables related to the pressure control system may contain oscillatory modes of response. In these cases, the decay ratio for each controlled mode of response must be less than or equal to 0.25. The temperature seasured by thermocouples on the discharge side of the valves shall return to within 100 F of the temperature recorded before the valve was opened. If pressure sensors are available, they shall return to their initial state upon valve Closure. During the 250 psig functional test, the steam flow through each relief valve as measured by the initial and final bypass valve (BPV) position shall not differ by more than 10 percent from the average relief valve steam flow as acasured by bypass valve position. During the rated pressure test, the steam flow through each relief valve as measured by change in NW(e) is not to differ by more than 0.5 percent of rated KW(e) from the average of all the valve responses. 3 24 3 Results During the heatup testing, all 15 SRVs were manually actuated. There was positive indication of steam dischargt upon actuation of each SRV. As each SRV was operated there was a sudden temperature rise on the SRV discharge tailpipe, the appropriate pressure Supple ent 9 P:ge 3 24-2 switch responded, and BPV position decreased to control reactor pressure. The Level 1 criteria was satisfied. All pertinent variables related to pressure control did not exhibit any oscillatory responses with decay ratios greater than 0.25. The SRV discharge line tempratures for five SRVs did not return to within 10 F0of L% temperature recorded prior to actuation as quickly as the other discharge lines; however, they did cool down sufficiently to indicate that the SRVs were not leaking. Shortly after the performance of this test a reactor scrau occurred and on the subsequent startup, the SRV tailpipe temperatures remained low, further verifying that the SRVs did properly reclose. Three SRVs had steam flow values, as measured by BPV position change, that differed from the average relief valve steam flow by greater ti.an 10%. The , bypass valve position was inadequate to get a proper value of steas flow from BPV position change. Upon the actuation of each SRV the BPV closed completely. Had there been more bypass steam flow, the EPV would not have closed completely and there would be a more L accurate value of SRV steam flow. This steau flow variance was reevaluated during the Test Condition Two SRV testing. All fifteen SRVs were manually actuated with the plant at rated pressure during Test Condition Two. Plant parameters related to pressure control were monitored on the GETARS computer, as well as other plant parameter responses, including generator load decreases. ; The Level 1 criteria was met based on three positive indications of steam discharge during the actuation of each valve. They were the sudden temperature rise in the discharge ta11 pipe, the positive indication of a NWe decrease during the valve actuations, and the response from the tailpipe pressure sensor of each valve being tested. The Level 2 criteria requiring that Pressure Control l Systen v=Mables did not exhibit any oscillatory , responses with decay ratios greater than 0.25, was , Supplement 9 Page 3024-3 : verified by the analysis of the GETARS data of the following variables: Pressure Regulator Output Control Valve Demand Control Valve #1 Position Narrow Ratige Pressure Generator Output (Gross NWe) GETARS data was also used to verify that the ctange in the plant's NWe following each SRV lift did not differ by more than 0.5% of the rated NWe from the average of all valves responses. All SRVs exhibited , a less than 5 5 MWe variation from the 68.$ NWe average variation, thus satisfying the Level 2 criteria. SRVs B21-F013J 0 and 821-F013N did not return to within 10 F of their initial tailpipe temperature values during the test. However, the temperatures 0 did return to within 10 F of their initial valuis when checked at a late time, thus satisfying a Level 2 criteria. Finally, part of the Licensing Consitzent 2.c.5 of the full power operating license was satisfied by this Test Condition Two relief valve test. It was demonstrated that all adjacent temperature readings were within 45 F of each other following a 10 second SRV lift with a suppression pool mixing system in operation. This concludes the relief valve testing to be performed during the Startup Test Phase Program. 1 I J i t l Suppiteent 9 Page 3 25-1 3 25 Turbine Stop Valve and Control Valve Fast Closure Trips 3 25.1 Purpose The purpose of this test is to demonstrate the response of the reactor and its control systems to protective trips in the turbine and generator. 3 25.2 criteria Level 1 For turbine / generator trips, there should be a delay of no more than 0.1 seconds following the beginning of control or stop valve olcsure before the beginning of bypass valve opening. The bypass , valves should be opened to a point corresponding to greater than or equal to 80 percent of their capacity within 0 3 seconds from the beginning of control or stop valve closure motion. Flooding of the main steam lines shall not occur following the turbine / generator trips. The positive change in vessel done pressure occurring within 30 seconds after either generater or turbine trip must not exceed the Levol 2 criteria by more than 25 psi. The positive change in simulated heat flux shall not. exceed the Level 2 criteria by more thaa 2 percent of rated value. Level 2 There shall be no MSIV closure in the first 3 minutes of the transient, and operator action shall not be required in that period to avoid the MSIV trip. ) The positive change in vessel dohe pressure and in j simulated heat flux that occur within the first 30 seconds after the initiation of either g3nerator or turbine trip aust not exceed the predicted values in the Transient Safety Analysis Design Report. For the turbine / generator trip within the bypass valves capacity, the reactor shall not scram for initial thermal pcder values less than or equal to 25 percent of rated. Supple ent 9 Page 3 25-2 l If the low-low set pressure relief logic functicns, f the open/close actions of the SRVs shall occur within 1 20 psi of their design setpoints. If any safety relief valves open, only onc valve may reopsn after the first blowdown. I 3 25 3 nesults During the Test Condition Two testing with a reactor power of 21.8%, a turbine / generator trip was initiated with a generator output of 151 MWe, by opening both generator output breakers CM and CF. I A rinctor scram did not occur following the tubine/ generator trip with the reat*or at 21.8% power. This is required at a reac W power < 25%, therefore, satisfying the Level 2 ceneria. l The East and West bypass valves began opening withic. 0.04 seconds and 0.06 seconds, respectively, following the be(*,nning of the cadtrol and stop valve closure. iiis satisfied the < 0.1 second ~ opening time required for the Level 1 criteria. j 1 The Level 1 criteria (applicable to Test Condition ) Six) requiring that the bypass valves open to a point corresponding to 3 80% of their capacity i within 0 3 seconds from the beginning of the control l and stop valves closure motions was not satisfied I during the Test Condition Two testing. The valves a-ly opened to 56.3% of their combined canac'ty at 0 3 seconds with the West Bypass Valve open 99 6%, and the East Bypass Valve open 12 7%. Repairs and off-line response time testing of the East Bypass l Valve Unitized Actuator were performed successfully Acing the MSR outage. The effects of steam flow on bypsss valve response time was further evaluatsd foldawing an inadvertent turbine trip from 50% reactor power on 7/20/87. The East and West Bypass Valves began opening within 0.025 seconds and 0.065 seconds, respectively, and had reached 3 80% of their capacity within 0.2 seconds which would have satisfied the above level I criteria if it had been applicable. Following the completion of Test Condition Five, with the reactor operating at 74.6% CTP with 73 6% core flow, an inadvertent turbine / generator trip / reactor scram was experienced. The date of this occurrence was 12/31/87. In accordance with Reference Number 3 in Section 1.5 of this report, an analysis of the data recorded during this event has Supplement 9 Page 3 25-3 been performed to determine if this %nadvertent trip can be substituted *or the full power r.urbine/ generator trip scheduled during Test Candition Six. This event occurred as a result of a Water Level 8 trip signal rather than from normal operating condition. Therefore, a code sianlation of a trip from full power and normal operating conditions was performed. It was concluded that all test criteria would have been satisfied if the test was performed as scheduled. Consequently, we have taken credit for this situation in accordance with j 10CFR50.59 and this change to the Startup Test Program has been submitted by letter NRC-88-0181, dated 7-14-88, as required by Fe.si 2 License Condition d.C(14). 1 1 4 J i 4 I Supplement 9 i P:ge 3 26-1 3 26 Shutdown from Outside the Control Room - 3 26.1 Purpose To demonstrate that the reactor can be brought from a normal, initial, steady-state power level to the hot shutdown condition and to verify that the plant has the potential for being safely cooled from hot shutdown to cold shutdown conditions from outside the control room. 3 26.2 criteria Level 1 l None Level 2 During the cold shutdown demonstration, the reactor ; aust be brought to the point where cooldown is , a initiated and under control. ! Durin6 the simulatt.d control room evacuation and hot shutdown demonstration, the resictor vessel pressure and water level are controlled using equipment and controls outside the control room. i I 3 26.3 Results 4 l During the simulated control room evacuation and hot : i shutdown test performed during Test Condition One, ] the designated Shutdown Crew, consisting of the 2 mininua shift complemerit, performed al activities ' associated with the reactor shutdown and control of the reactor vessel water level and pressure from outside the Control Room. The reactor vessel pressure and water level were l controlled for a period of over thirty minutes [ following successful reactor shutdown and isolation from outside the Control Roon by the mininua shift ; complement, which successfully meets all test ! criteria and performance objectives of the : applicable governing documents. I [ i , i i 1 i Supplement 9 1 Page 3 26-2 i The test sequence of events was as follows: Time Event 1223 Test Start Time (Hi Come Announcement) 1224 "Shutdown Crew" Evacuation of Control Roon l l 17.24 APRMs A&B to Standby (to initiate Reactor [ i Scram) 1224 Relay TTR-2 manually tripped (to initiate Main Turbine Trip) ! 1225 Main Steam Line Radiation Monitors to Standby ( initiate MSIV ! solation) i 1226 Restoration of APRMs A&B and the Main Steam l Line Radiation Monitors to the Operate l positions 1226 Exit Relay Roon ; 1228 Transfer Switches operated at Remote i 2 Shutdown Panel (RSP) (RSP Control) , 1230 RHRSW started at Remote Shutdown Panel (RHR } Service Water Pumps A&C) ; 4 1233 RNR Pump A started at Recote Shutdown Panel j l 1233 Div II Transrer switch operated (Div II D.C. t ESF Power) ] 4 1234 RCIC initiated from Remote Shutdown Panel i ) l 4 4 1235 RcIC at rated flow (600 spa) ! i 1237 "A" SRV cycled from Remote Shutdown Panel , (0 pen for approximately seven seconds) 1238 "B" SRV cycled from Remote Shutdown Panel i (Ocen for appt nimately nine seconds) 1239 start of stable Control
- iriod in Hot l Shutdown }
1313 Completion of stable Control Period in Hot Shutdown i i l l ! l ! l 1 i ! Supplement 9 r Page 3 26-3 , r Time Event 1313 Transfer Swi+.ches operated (RSP Transfer to Control Roon Control) 1313 Te:t Termination The remaining testing within this section, involving i a demonstration of the plant's capability to reach ! cold shutdown conditions from et,tside the control ;
- room, is scheduled to be perfened following the ;
MSIT simultaneous full closure test in Test d Condition Six. : A i + t 9 1 1 i j i i : i l j
- i t
. I 1 l 1 i h i i I I i t a i l 1 i i I 2 i l i. l I t Supplemont 9 ! Page 3027-1 i 3 27 Flow Control
- 3 27.1 Purpose ,
- a. To dstermine the correct gain settings for the individual recirculation controllers.
- b. To demonstrate plant response to changes in recirculation flow in both local manual and master nahual mode,
- c. To set the limits of range of operation for the ,
recirculation pumps. 3 27 2 Criteria Level 1 I t The Oransient response of any variable related to
- the recirculation system to any test input must not
- diverge. !
Level 2
- The decay ratio of the speed loop response shall be 10.25 at any speed. ;
riow control systen limit cycles (if any) must produce a turbine steam flow variation no larger , J than 10 5 percent of the rated steam flow value. 1 ! l' The APRM neutron flur trip avoidance sargin shall be 37.5 percent, and the heat flux trip avoidance ; margin shall be 35.0 percent as a result of the , recirculation flow control maneuvers. l L i I 3 27 3 R alts ( , i In Test Condition Two, g 45 step change testing was performed on both recirculation systen speed control ; loops in the local manual mode at 38.8% Reactor ! power and 47.55 core flow. ; r A review of the data recorded indicates no variables related to the recirculation system were divergent. A qualitative review of the speed response of the A Reactor Recirculation MG Set verified that the decay rgtio was < 0.25 for the g 45 speed steps j performed. , t ) ! i ) d Supplement 9 f P:32 3 27-2 L The B Reactor Recirculation (M) MG 50t exhibited t ' limit cycle of approminately 21/25 speed peak-to-peak when operating at 38% speed. Due to i this limit cycle, the "B" speed loop response Decay l Ratio could not ba verified and will be ratested when controller optialzation is performed in Test Condition Three, f Flow contrsi systen limit cycles were verified and the peak-to-peak changi in gross generator output , during steady state conditions was less than + 0.55 t of rated generator output or 11.5 We peak to-peak. His criteria was patisfied with the largest . observed generator output limit cycle of 10 55 W e . peak-to-peak (+ .46% of rated output). he peak APRM neutron flus was 57.715. This APM r l reading includes an APRM gain adjustment factor of
- 1.25 which was required due to a high core peaking :
! factor. The calculated APRM neutron flus trip ; ] avoidance margin was 60.295, satisfying the 3 7 5% : criteria. 1
- The ninlaus heat flus trip avoidance margin was 22 395 for the increasing speed steps, satisfying t j the criteria of 3 5.05. ,
) t i In Test Condition tree, testing was performed to i i demonstrate that the plant response to changes in i j recirculation flow was stable following flow control systen tuning. Initial settings were also input to , ! the dual limiter portion of the Master Flow - Controller control circuitry in the Master Manual i Mode of recirculation flow control. .let pump , l baseline data for compliance with Technical ! I 5pecifica',:!ons Surveillance 4.4.1.2, Jet Pumps Operability, along the 755 rod line was obtained, a , The test wa3 conducted in three sessents to support ; the above objectives. The first segment consisted of individual l Recirculation MG Set + 9111 speed step tests in , i local manual mode at The applicable region of r highest gain for each Recirculation MG 5et. The !
- speeds associated with these regions of high gain l were previously identified as 655 for RR MG Set "A" ;
i and 691 for RR MG Set "8". The second segment i consisted of + 4 6% speed steps in the Master Manual Mode of operaiion at RR MG Set speed / flow points of l 961 flow, 70% RR MG Set speed and 50% RR MG Set ; I speed. ; d t i l J ! i l Supplement 9 P:ge 3 27-3 The third segment consisted of adjustment of the dual limiter portion of the Master Flow Controller and obtaining baseline jet pump d/p dsta at 2% speed steps between 61% to 955 core flow along the 75% rod lina. The sequence of testing was to reduce the Master Flow Controller H/A station output slowly until no further core flow decrease occurred. This point was then labelled as the minimum core flow point and loop speeds, core flow and "lo-pot" setting of the dual limiter were recorpod. The point was verified by Master Flow Controller H/A station output increases untti the core flow just started to respond. The Master Flow Controller was then adjusted in 2% increasing speed steps with flow allowed to reach steady sta'se for each step. Jet pump baseline data was recorded when steady state conditions were obtained at each new speed plateau. This increase was halted at the electrical speed stop on RR MG Set "A". The "hi-pot" setting of the dual limiter for the Kaster Flow Controller was then adjusted to slightly lower the RR HG Set speeds to be less than the electrical high speed stop setting. Tnis point was then labelled as the mariaua core flow point and loop speeds, core flow and "hi-pot" setting of the dual limiter were recorded. The dual limiter setpoints at the 755 rod line are tabulated below: Minimun Flow Datat Maximun Flew Data Core Flow: 611 Core Flows - 955 HG Set "A" Speed t 461 _ HG Set "A" Speedt 78.5% HG Set "B" Speed: 45.51 HG Set "B" Speed: 78 1% "Lo-Pot" Setting: 1 "Hi-Pota Setting: _68 These settings will be evaluated again during the Step Change TestinF/Rinp Test seneduled during Test Condition Six. A review of the STARTREC traces for the above testing indleated thtt no variables related to the Recirculation Systes were divergent, thereby satisfying the Level I criteria for this test. ~. i l Supplement 9 Page 3027-4 A qualitative review of the applicable STARTREC traces showed that the speed Icop resportse of "A" RR MG Set loop controller, "B" RR NG Set loop controller and the Master Flow controller to the required steps all met the Level 2 < 0.25 damping critaria. All of the post step steady state turbine steam flow variations were less than 11 5 MW(e) peak to peak, satisfying the applicable Level 2 ciiteria. The results are tabulated below Meets Criterir. W(e) of <11.5 MW(e) Step Variation Peak _to Peak Loop A 9-11% decrease 5,31 Yes (655 speed) Loop A 9-115 increase 7.45 Yes (655 speed) Loop B 9-11% decrease 6.83 Yes (695 speed) Loop B 9-115 increase 8.07 Yes (695 speed) Both Loopc 4 6% decrease (96% flow) 6.83 Yes Both Loops 4-65 increase (96% flow) 7.45 Yes Both Loops 4-65 decrease (TOS speed) 6.83 Yes Both Loops 4-65 increase (705 speed) 7.45 Yes l 1 I l Supplement 9 : Page 3027-5 l l l Meets Criteria MW(e) of <11 5 Mk(e) Step Variation Peak to Peak j l Both Loops l 4-65 decrease I (50% speed) 8.69 Yes ; Both Loops 4-65 increase (50% speed) 6.83 Yes . I ! All seras avoidance margins were met as shown by the l l values tabulated belows i l APRM High Neat Flus ' l Flus Margin Trip Avoidance Meets Step Margin 37 55 Margin 15.0% Criteria Loop A 9-115 increase 38.1% 36.1% Yes j (655 speed) Loop B 9-11% increase 37.0% 34.9% Yes ! (695 speed) Both Loops 4-65 increase j (96% flow) 37 1% 35.4% Yes l Both Loops ; 4-65 increase (705 speed) 41.6% 28.2% Yes ; l Both Loops ! 4-65 increase , (50% speed) 52.6% 25.75 Yes With respect to the previously noted limit cycle (2 ; i 1/25 speed peak-to-peak) on the B Reactor Recirculacion MG Set speed loop at 381 speed, this problem has since disappeared htter Scoop Tube can shaping and has been attributed to an anomaly of the , original cas. : Both A and B MG Sets do, however, exhibit speed , oscillations of approximately 35 speed peak.to-peak l in the 24-281 and 52-565 speed ranges. Yhese ; oscillations are due to an inherent anomaly in the [ fluid coupler hydraulic systes and this phenomenon i is not specific to Fermi 2 MG Sets. The l t \ l I i Supplement 9 Page 3 27-6 oscillations in the 24-28% speed range has been overcome by prohibiting operations below 28% speed. , Investigation as to the effect of the oscillation in the 52-565 soe64 region was conducted and the effect of these oscillations have been minimized by unbalancing the Reactor Recirculation MG Set speeds within the limitation of Tech. Spec. 3 4.13 while ramping flow upward through this r6gion. In Test Condition Sir, step change testing in bJth ; local manual and master manual wn< performed on the 1001 Load Line to demonstrate that the plant response to changes in recirculation flow was stable. Individual Recirculation MG Set + 4-65 speed steps in local manual mode were performed at the previously identified highest gain region speeds of 65% for Reactor Recirculation MG Set "A" and 695 for Reactor Recirculation MG Set "B". Master manual mode speed steps of + ~ 2-31 were performed at Reactor Recirculation MG Set speed / flow points of 100% flow, [ 75% Reactor Recirculation MG Set speed and 655 > Reactor Recirculation MG Set speed. A qualitative review of the applicable STARTREC traces showed t. hat the response of "A" RR MG 5et Speed Loop Controller, "B" RR MG Set 5;eed toop Controller and the Master Flow Controller to the required steps all met the Level 1 criteria for non divergence and the Level 2 < 0.25 damping criteria. , All of the post step steady state turbine steam flov variations were less the 11.5 W(e) peak to peak, satisfying the applicable Level 2 criteria. The results are tabulated below Meets Criteria NW(e) of <11.5 W(e) Step Variation Peak to Feak Loon A 4-6) decrease 7.4 Yes (655 speed) Loop A 4-65 increase 6.8 Yes (655 speed) Loop B 4-65 decrease 93 Yes ] (695 speed) t Supplement 9 Pag] 3 27-7 Meets Criteria W(e) cf <115 W(e) Step Variation Peak to Peak Loop B 4-f,5 increase 6.8 Yes (695 speed) Both Loops 2-35 decrease (100% flow) 6.2 Yes Both Loops 2-31 increase (100% flow) 6.2 Yes Both Loops 2-3% decrease (75% speed) 5.6 Yes Both Loops 2-35 increase (755 spsed) 7.4 Yes Both Loops 2-35 decrease (655 speed) 5.6 Yes Both Loops 2-35 increase (655 speed) 8.1 Yes l l All scras avoidance margins were met as shown by the l values tabulated below: APRM High Heat Flux Flux Margin Trip Avoidance Meets l Step Margin 17.51 Margin 15.01 Criteria Loop A 4-65 increase 26.9% 20.2% Yes (655 speed) Loop B 4-65 increase 22.75 21.61 Yes (695 speed) Both Loops 2-31 increase (100% flow) 12.1% 10.45 Yes i [. r Supplement 9 I P ge 3 27-8 I t APRM High Heat Flux ! Flux Margin Trip Avoidance Meets i Step Margin 37.55 Margin 15.05 Criteria : i Both Loops f 2-35 increase l (755 speed) 19.25 20.25 Yes , ! i Soth Loops l 2-31 increase i (655 speed) 28.05 19 75 Yes j The remaining testing in this section to obtain Jet i " Pump baseline data along the tool Rod Line and the verification / adjustment of the dual limiter t setpoints of the master flow controller has not yet 1 been completed. : i i I 'l I l i i l i ! i ! t k 1 ; I [ ] ! r t 4 4 i' i l ' i - . . - - - - , -t Supplement 9 $ Pag) 3 28-1 3 28 Recirculation syat< 3 28.1 purpose
- a. To verify that the feedwater control system can satisfactorily control the water level without a resulting turbine trip / scram and obtain actual pump speed / flow,
- b. To verify recirculation pump startup under pressurized reactor conditions.
- c. To obtain recirculation systes performance data.
- d. To verify that no recirculation system t cavitation occurs in the operable region of the power-f2cw asp.
3 28.2 criteria i Level 1 The response of any level-related variabita during pump trips must not diverge. l l Level 2 l The simulated heat flux margin to avoid a scraa shall be greater than or equal to 5.0 percent during i the c.ie pump trip recovery. The APRM nargin to avoid a scraa shall be greater I
- than or equal to 7.5 percent during the one pump l trip recovery, r l
Duririg the noncavitation verification, runback logic I ). shall have settings adequate to prevent operation its ! areas of potential cavit4 tion. l During the one pump trip, the reactor water level margin to avoid a high-level trip (L8) shall be " greater than or equal to 3 0 inches. l 3 28 3 Results [ Dur.og Test Condition Two, recirculation system I baseline performance data was recorded at 38.8% l reactor power and 47.5% core flow and at 48% reactor : power and 55.7% core flow. l i I f l i Supplement 9 Page 3028-2 , Baseline Recirculat' ion System Performance data at Test Condition three power - flow conditions was collected at 47% power and 100% core flow. Also during Test Condition Three, n test was run to verify that the recirculation pump runback limits are sufficient as to prevent operation where recirculation pump or jet pump cavitation is predicted to occur. The test was conducted by establishing total core flow at 905 (+ 3%) of rated at a reactor poser of 44.2%. Both Recirculation MC Set Scoop Tubes were locked and while reducing reactor power by the insertion of control rods, jet pump dp, recirculation pump vibration, drive flow, pump dalta pressure, and pump suction temperatures were continuously monitored for indications of pump cavitation. Throughout the power reduction to the actuation of Ll miter #1 at 23 6% of rated feedwater flow and 27.4% of rated reactor power, no indications of pump cavitation were observed. Reactor power was further reduced to 21.7% rated feedwater flod and 25 3% of rector power at which point the power reduction was stopped due to indication of an increasing width of the recording of reactor core delta P which could be an early indication of cavitation. R erefore, it may be concluded that the runback logic settings are conservatu ely adjusted such that operation in areas of potential cavitation is prevented and that the Level 2 criteria has been satisfacterily met. During Test Condition Six, recirculation systen baseline performance data was recorded at 98.6% CTP and 100% CF. Also during Test Condition Six, the one recirculation pump trip test from 95.9% CTP and 975 CF was conducted to verify that the feedwater control system can satisfactorily control the reactor water level without a resulting turbine trip /scras and to obtain data for actual recirculation MG Set speed vs flow in the str41e recirculation loop configuration. The test was conducted from the above initial conditions by placing the Reactor Recirculation H3 Set "A" notor CMC switch to the Off position. l Supplement 9 P:ge 3 28-3 [ Followleu the transient and after stable conditions were reached, the "B" Reactor Recirculation E Set i speed was reduced to 75% and speed / flow data in the single loop configuration was gathered in 2% (+ 1%) decreasing increments until 305 speed was reached. Test Condition Four was then entered by placing the "B" Reactor Recirculation MG Set motor CMC switch to the Off position. Upon completion of the required Test Condition Four tests (as described elsewhere in , this report), "A" Reactor Recirculation MG Set was ; restarted to exit from Test Condition Four. Upon attempting the restart of "B" Reactor Recirculation MG Set, its discharge valve 531 F031s failed to [ close, and therefore restart logic could not be ; c d h fied. We plant was shut down to effect n n 's to this valve and after restart and power l myton to the same approminate initial conditions of 59 9% CTP and 46.9% CF, "B" Reactor Recirculation t MG 5et was tripped and a restart attempted with the ; same resulting failure of s31-Fo31s to close, ne , d plant is currently shut down to further investigat. [ the probles. 3 t A qualitative review of t: e wnse of level i related variables recorded t. is the one pump trip . from 95 95 CTP showed that the Level 1 criteris for t this test was set since none were divergent. The
- margin to the reactor water high level (L8) trip was ,
1 5 1 inches thereby satisfying the Level 2 criteria ! of 3 3 0 inches. 1 The Level 2 criteria of APRM and Simulated Heat Flu ! Margins to Scran could not be verified for the one pump trip recovery due to the inability to restart , the "B" Reactor Recirculation MG Set. \ \ ' he restart of the "A" Reactor Recirculation NG Set from Test Condition Four was, however, evaluated to . this criteria with the APRM Margin to Scras being ! 79.65 and the Simulated Heat Flum Margin to Scran , being 22.4% thereby satisfying the Level 2 criteria. i-he balance of this test to verify the restart of recirculation pumps under pressurized reactor l conditions with the other pump running will be l l performed when the plant returns to operation. I l 1, . t I , i Supplement 9 P2ge 3 29-1 3 29 Loss of Turbine-Generator and Offsite Power 3 29.1 Purpose
- a. To determine the reactor transient performance during the loss of the main generator and all offsite power.
- b. To demonstrate acceptable performance of the station electrical supply system.
3 29.2 Criteria Level 1 The reactor protection system, the diesel-generator, RCIC and RPCI must function properly without manual assistance. HPCI and/or RCIC system action, if necessary, shall keep the reactor water level above the initiation level of low-pressure core spray, LPCI, and automatic depressurization systems. Level 2 If the low-low set pressure relief logic functions, the open/close actions of the SRVs shall occur within +20 psi of their design setpoints. If any safety relief valves open, only one may veopen after the first blowdown. 3 29 3 Results The test was initiated during Test Condition Two by isolating the plant from off-site power by simultaneously opening both the 345 KV and 13 2 KV feeds to the in-plant busses. It was demonstrated that the following actions occurred once the test was initiated without any operator assistance:
- 1. The Reactor Protection System automatically scrammed the reactor.
- 2. The Turbino/ Generator Protection System automatically initiated a trip and fast closure of the Main Turbine steam admission valves.
3 The Emergency Diesel Generators automatically started and properly loaded the ESF busses, and SupplcEent 9 Pag 3 3 29-2 ; l
- 4. Control of reactor water level and pressure during transient conditions were uaintained.
It was also demonstrat(d that the required equipment and support systems operated satisfactorily without dependence on off-site power sources for the extended test duration of 30 minutes. No automatic initiation signal /setpoint was received for either (IPCI or RCIC. The lowest reactor water level reached during the test was 138.8 inches. The Level 1 setpoint of 31.8 inches, at which Core Spray, LPCI and ADS are initiated, was therefore avoided by a significant margin. Based on the above, the Level I criteria for this test was successfully set. Following the first blowdown, only SRV B21-F013A reopened. This satisfies the Level 2 criteria requirement that specifies only one SRV may open at that time. The low-low set pressure relief function for two low-low set valves, SRV "A" and SRV "G" was actuated during the test. On increasing reactor pressure, six SRVs lifted at a pressure of 1100.1 psi. These actuations were in accordance with the Level 2 criteria required for this test. This concludes all Loss of Turbine / Generator and Off-Site Power testing during the Startup Test Phase program. l 1 Supples nt 9 l Piga 3 30-1 ! 3 30 steady-state vibration 3 30.1 Purpose To determine the vibration characteristics of the primary pressure boundary piping (NSSS) and ESF (ECCS) piping systems for vibrations induced by recirculation flows, hot two-phase forces, and hot hydrodynamic transients; and to demonstrate that flow-induced vibrations, similar in nature to those expected during normal and abnormal operation, will not cause damage and excessive pipe movement and vibration. 3 30.2 criteria Level 1 The measured vibration levels of the piping shall not exceed the acceptable specified values. Level 2 The seasured vibration levels of the piping must not exceed the expected specified values. 3 30 3 Results During Test Condition One, the RCIC Steam Supply Line inside the drywell and tne RCIC Pump Discharge Line near its connection to the Feedwater Line were monitored for vibration using installed sensors during a vessel injection at rated conditions. Evaluation of the data showed that all vibration levels were within acceptable values. During Test Condition Two, steady state vibration ' was seasured for selected piping systems at 25% (t 5%) of rated steam flow and at 50% (1 5%) of rated core flow. Data was initially gathered for seven piping systems consisting of Feedwater, Main Steam, Reactor Recirculation, RHR, SRVs D&J, HPCI and RCIC. More data was collected at a later date for eight locations on the Main Steam piping and one location on the RCIC piping at 25% and 29% rated steam flow. This extra testing was necessary because the Level 1 criterion for six of these locations were exceeded in the initial set of data. Also, more data was needed to determine the impact of the removal of j Supplement 9 Paga 3 30-2 anubbers from piping between the Turbine Control Valves and the High Pressure Turbine. A total of eight Level 1 criterions for instruments D-015, D-016, D-017, A-014, A-015, and A-016, were exceeded in this second set of data. However, based on hand held vibration acasurements and/or detailed pipe stress analysis by Sargent and Lundy, all criteria violations were found acceptable. Revised criteria levels for selected sensor locations were incorporated into future test plans. During Test Condition Three, vibration data was collected to determine the flow induced vibration responses of the Main Steam Lines, Reactor Recirculation Loops, Feedwater, HPCI, RCIC, RHR and Safety Relief Valve piping during steady-state vibration hardwired testing. Steady-state vibration data was obtained and analyzed for 80 (+ 5)5 and 100 (+ 5)% of rated core flow, and 50% (1 5%) and 75% (I 55) of rated steam flow. Post transient steady-state data was also obtained following two HPCI RPV injection for HPCI piping sensors. Dere was a total of two (2) exceedences to the Level 1 criteria as follows during the 80% core flow data collection: Level 1 Measurement S,ensor e mils p-p mi]s p-p, A-014 10 11.3 A-015 14 49.6 For sensors A-014 and A-015, it was determined that their readings were unreliable, and that vibration for this area of piping is acceptable based on the readings of sensors D-009, D-010 and D-011. D ere was one exceedence to the Level 2 criteria during the 100% Core Flow data collection. Level 2 Measurement Sensor inch p-p inch p-p SA-RZ 0.024 0.027 r i Supple:2nt 9 Pega 3 30-3 The Level 2 criteria exceedence for sensor SA-h2 was evaluated and considered to be acceptable. Review of the same sensor data at 25% steam flow and 50% core flow showed satisfactory peak-to-peak amplitude. There was one exceedance to the Level I criteria during the 75% Steam Flow data collection. Level 1 Neasurement Sensor inch p-p inch p-p A-006 50 77 For sensor A-006, it was determined after an evaluation by Sargent and Lundy given the as-built location of the sensor, reviewing the design calculations and similar test data, that the vibration criteria of A-006 would be acceptable to a new increased vclue of 84 mils peak-to-peak. Although sensor A-601 (RHR Head Spray Line) appeared to have exceeded its Level 1 criteria based on vibration data taken at 70.1% steam flow, it had previously been determined from data at approximately 60% steam flow that this sensor was not giving a true reading. This diLgnostic data was collected to determine sensor operability prior to raising power and it was noted that this particular sensor's reading was not consistent with the other eight (8) neighboring sensors on the same line. An evaluation was performed at that time by Sargent and Lundy and it was determined that this sensor reading was not correct and would not be a restraint to raising reactor power. There was also one exceedance to the Level 2 criteria during the 75% Steam Flow data collection. Level 2 Heasurement Sensor inch p-p inch p-p SA-RZ 0.024 0.026 The Level 2 criteria exceedance for sensor SA-RZ was previously evaluated during the 100% Core Flow data collection and found to be acceptable. There were no piping vibration criteria exceedances during the 50% (+ 5%) Steam Flow data collection. t Supplement 9 Page 3030-4 Post transient steady-state data following the HPCI RPV injection was analyzed and found acceptable; however, this data collection was repeated due to the subsequent repiccement of E41-F005, HPCI Discharge Check Valve. Results from this additional data collection during a HPCI RPV Injection in October of 1987 were also acceptable. During Test Condition Six, vibration data was collected to determine the flow induced vibration responses of the Main Steam Lines, Reactor Recirculation Loops, Feedwater, RCIC, RPCI, RHR and Safety / Relief Valve piping during steady state operation of the plant with Main Steam flow between 95 and 100% of rated steaa flow. Vibration data was collected during the following steady-state conditions: Date Performed Power Core Flow Main Steam Flow July 9, 1988 97 1 97 95 3 July 11, 1988 96.74 97 5 95 3 In the first data set, there were two (2) apparent Level 1 criteria exceedances associated with accelerometers A-052 and A-053 These accelerometers are located on Safety Relief Valve B21-F013E pip!ng. However, based on Sargent and Lundy recalculations of allowables transmitted to the site on May 20, 1988, the Level 1 criteria associated with these sensors were being revised and the sensor readings were within the new Level 1 criteria. Level 1 Measurement New Level 1 Sensor Mils p-p Mils p-p Mils p-p A-052 9 16 17 A-053 8 12 16 There were no Level 2 criteria exceedances. Supplcc nt 9 Pcge 3 30-5 In the second set of data performed to the revised procedure, no Level 1 and only one (1) Level 2 criteria violation were found. Level 1 Level 2 Measurement Allowable Expacted Sensor Mils p-p Mils p-p Mils p-p SB-RZ2 76 110 56 This exceodance has been reviewed by General Electric and has been found acceptable. Also during the Test Condition Six steady state testing, the vibration traces corresponding to Lanyard Potentiometers RA-SX1 (Recirculation Loop A), and RB-HX4 (Recirculation Loop B) and Accelerometer A-108 were indicative of bad sensors. Based on other adequate data from nearby sensors, the omission of these three from the data set was found acceptable by both Sargent and Lundy and General Electric. --- - _ . . - - , - - _ - - - ----_.,_.._m_,, _ , _ , , - __.__m._m.__ _ __.. _ - - , _ , , - - .--,____-_m-,- ,,,., -. Supplc ent 9 Page 3 31-1 3 31 Recirculation Systen Flow Calibration 3 31.1 Purpose To perform a complete calibration of che installed recirculation system flow instrumentation. 3 31.2 criteria Level 1 None Level 2 Jet pump flow instrumentation is adjusted so that the jet pump total flow recorder provides a correct core flow indication at rated conditions. The APRH/RBH flow-bias instrumentation is adjusted to function properly at rated conditions. The flow control system shall be adjusted to limit maximum core flow to 102.5 percent of rated flow by limiting HG set scoop tube position. 3 31 3 Results During Test Condition Three at a reactor power of 455, a total core flow calibration was performed using Reactor Engineering procedure 56.000.02. This was the first core flow calibration perforned and therefore, approximately 55 margin : ras established between rated and indicated core flow. During the initial run, the jet pump milli-volt readings were found to be varying making it difficult to obtain accurate readings. Several readings were taken at each square rooter. He highest and lowest readings were averaged together and the average value was recorded. The Reactor Engineering procedure required that the jet pump square rooter output be within ,25 na of the expected output based upon measured input. Bis requirement was not initially met. The Reactor Engineering in-house code calculated a total core flow of 97.7%. This value compared well against the General Electric code, JRPUNP, (which calculated core flow to be 97.6%). This is a very good agreemera since the Reactor Engineering code used jet pump instrument span from I&C calibration sheets, while JRPUMP used the design instrument span Supplerent 9 Page 3 31-2 of 10-50 ma. An RC network :as developed to filter the jet pump silli-volt readings and the Reactor Engineering procedure was run a second time. Milli-volt readings were taken simultaneously from the input and output jacks of the square root extractors. This method enabled us to meet the requirement that the output of the square root be within .25 na of the expected output. The filter helped, but did not prevent, the oscillations in the milli-volt readings. The Reactor Engineering procedure was run a third time using a different filter with a 4-5 second time constant. The milli-volt readings were still unstable but average values were recorded. Core flow was calculated to be 100.0% by the Recctor Engineering code, while JRPUMP calculated core flow to be 99.8%. The flow calibration was completed by adjusting B21-602 A, B and B31-607 A, B, C, D summers, which satisfies the Level 2 criteria for the adjustment of instrumentation providing core flow indication and APRM/RBM flow-bias. The recirculation system was placed in HASTER MANUAL. Speed and flow data was collected while flow was decreased from 100% to 80%. Flow vs speed data was plotted for this range. This data was extrapolated out to 102.5% core flow to obtain the corresponding speed. Flow was increased to 95%. The recirculation system was placed in the LOCAL MANUAL mode. MG Set "A" speed was increased to 850 rpm (equivalent to 102.5% flow). The mechanical stop was set at this speed. The electrical stop wrs set 6/64" before the mechanical stop based upon the scoop tube positioner. This position is at S40 rpm (equivalent to 101% flow). The mechanical and electrical stops were set in a similar canner on MG Set "B". MC Set "A" speed was reduced and MG Set "B" was increased. The mechanical stop was set at 868 rps (equivalent to 102.5% flow). The electrical stop was set at 855 rpm (equivalent to 100 7% flow). This is 14/64" before the mechanical stop based upon the scoop tube positioner. This satisfies the Level 2 criteria of limiting the maximum core flow to 102.5% of rated by limiting the MG Set Scoop Tube positions. Later during Test Condition Three at a reactor power of 71% and 94.5% core flow, a total core flow calibration was performed using Reactor Engineering ' Procedure 56.000.02. This power and core flow were Supplem:nt 9 P:gs 3 31-3 sufficiently high to obtain data at the upper end of the Test Condition Three window. The latest revision of the Reactor Engineering Procedure incorporated the lessons learned from the previous core flow calibration performed at 45% CTP. The RC filter network was modified for ease of connection and proper resistor configuration and the procedure was modified to ensure that this filter was properly applied. This resolved previous concerns with APRM flow unit GAFs and jet pump summer adjustments and facilitated the adjustment portion of this test. The basic sequence was that data was gathered following a calibration check of jet pump loop instrumentation. The data was used in the in-house Reactor Entineering code to calculate total care flow, jet pump loop flows and APRM Flow Unit Gain Adjustement Factors. Total core flow correction was not required. Jet pump loop flows required adjustments as one loop was indicating greater than calculeted flow and the other was indicating less than calculated flow. The APRM flow units also required adjustment because of the increased drive flow required for the same core flow (decreased M-ratio). These adjustaents were calculated as Gain Adjustment Factors (GAFs) and were successfully applied to the flow units. The existing settings of the Reactor Recirculation MG Sets scoop tube high speed electrical stops required that a small adjustment ta made to the high speed stops so that the required speed to obtain 100% indicated flow could be reached. This limited the available data for extrapolation of the final 102.5% flow equivalent speed and coupled with the requirements for conservatism with respect to Technical Specifications setpoints combined to provide sufficient error in the settings of the electrical high speed stops to preclude obtaining 100% flow at predicted high speed stop settings. The procedure allows for readjustment of the high speed stops if required to obtain 100% core flow, but does not require the adjustment. Due to the nonlinear behavior of two phase flow losses from the 50% rod line to the 75% rod line, it was determined that further adjustmen h would not be done at this time, but rather would be adjusted at the 100% rod line at approximately 90% power. I ,_ . - _ _ , _ _ _ , . __ ~ , _ . _ _ __ ._,_.,.,__m_. Supple:;nt 9 P;ge 3 31-4 The Level 2 criteria associated with this test were satisfied as follows:
- a. The B21-K602A/B jet pump flow loop sunsers were adjusted in accordance with the flow calculated in Reactor Engineering Procedure 56.000.02.
The computer code output indicated the following gain adjustment factors for B21-K602A/B: B21-K602A (GAF) = 0 9520 B21-K60?B (GAF) = 1.0380 Composite gains were calculated per Reactor Engineering Procedure 56.000.02 and entered into the Reactor Engineering data book. The composite gains for the jet pump loops were: B21-K602A (CGAF) = 1.000 B21-K602B (CGAF) = 1.040
- b. The B31-K607 A,B,C,D summers ( APRM/RBM flow-bias) were adjusted using the GAFs calculated in Reactor Engineering Procedure 56.000.02.
The computer code output indicated the following Gain Adjustment Factors for flow units B31-K607A through B31-K607D: Flow Unit A - B31-K607A (GAF) = 1.021 Flow Unit B - B31-K607B (GAF) = 1.039 Flow Unit C - B31-K607C (GAF) = 1.034 Flow Unit D - B31-K607D (GAF) = 1.019
- c. The mechanical high speed stops were set at 102.5% core flow which equated to 81.4% speed (912 rps) for RR MG Set A and 82 35 (922 rpm) for RR MG Set B. The electrical high speed stops were set at 100% core flow for the "A" MG Set which equated to 80.0% speed (896 rpm), and 100% for the "B" HG Set which equated to 80.5%
speed (902 rpm). , After completion of the 100 hour commercial run at > 905 CTP but prior to entry into Test Condition Six, another c&libration of the recirculation systen flow instrumentation was run at 95 5% CTP to determine the scoop tube electrical and mechanical stop positions. I Supples:nt ? Page 3 31-5 Reactor Zugineering Procedure 56.000.02 was used to determine the actual core flow at the time of this test. The Process Computer program OD-3 option 2 was run to obtain the initial conditions for this test, and based on that OD-3 Option 2 edit, the core flow was 96.21 M1ba/hr. Th.a core flow as indicated on the control room recorder B21-n613 was 94 Mlbs/hr. However, the actual core flow waa determined to be 91.82 Mlbe/hr and the jet pump summers and flow units were adjusted to bring the indicated flow within acceptable agreement (within + 2% of rated flow) of the calculated flow. After the adjustments were made, the indicated core flow on the control room recorder B21-R613 was 91.0 Mlba/hr and the core flow obtained via the Process Computer was 92 77 Mlba/hr. Rcactor Engineering Procedure 56.000.02 was also used to determine the Gain Adjustment Factors to be applied to the recirculation system flow units which were used to adjust the Technical Specification regiired flow biased rod block and scram setpoints. ?rior to these adjustments, the flow biased APRM rod blocks were set approximately as the rated load line causing rod blocks when approaching the rated load line. The average GAF applied to the flow units was 1.09 which raised the APRM flow biased rod block line about 6% above the rated load line. This provided the margin to be able to operate at or near the rated load line with the rod block annunciator cleared. The calculation portion of REP 56.000.02 was run twice: first by using an estimated number of 33 0 Hlbm/hr for the rated recirculation drive flow and the second time by using the actual value of the rated drive flow as determined from the new H-ratio. The new M-ratio was determined to be 2.198 for loop A and 2.168 for loop B. The design M-ratio is 1.84 and the difference in the design versus actual M-ratios suggest that at the present time, the actual rated recirculation drive flow is 31.4 Mlbm/hr and not 35.2 Mlbm/hr. An evaluation run of REP 56.000.02 will be performed later at rated plant conditions to obtain the updated M-ratio, core dP and indicated versus actual core flow values. Once the core flow and recirculation flow unit adjustments were made, core flow was reduced from 91 Hlbm/hr indicated (92 93 M1bm/hr from the Process Computer) to 81 Hlbm/hr indicated (83 28 Mlba/hr from the Process Computer) in approximately 2% core flow steps. Core flow versus recirculation MG Set speed data was collected at each step. 1 Suppleutnt 9 Paga 3 31-6 The core flow versus percent of rated MG Set speed was graphed and used to extrapolate the Recirculation MG Set speed corresponding to 102 5% of rated core flow. The extrapolated speed was determined using the core flow signal from the Process Computer instead of Control Room recorder B31-R613 to set the mechanical stops as this resulted in a more conservative (lower) Recirculation MG Set speed. Or.ce the MG Set speeds corresponding to 102 5% core flow was determined, the mechanical and electrical stops for both A and B MG Sets were set. This was accomplished by first increasing the recirculation speed to 79% in individual manual mode. At this point, A MG Set speed was increased to 86.2% speed (equivalent to 102.0% core flow) using the M/A station B31-R621A. The previous electrical and mechanical stop settings were adjusted to allow this speed increase. The "A" MG Set scoop tube was then lockeu and the speed manually cranked up to 970 rpa (86.6% speed) corresponding to 102.5% core flow and the mechanical stops set. The MG Set A speed was manually cranked down to 953 rpa (85.1%) to set the electrical stops. This was the speed associated with 100 9% of rated core flow. Once this was accomplished, the A M" Set scoop tube was unlocked and the speed reduced to 79% speed to match the B MG Set speed. The same procedure was repeated for B MG Set, and the mechanical stop was set at 960 rps (85 7%) corresponding to 102.2% core flow and the electrical stop was set at 948 rpm (84.6%) which is equivalent to 101.0% of rated core flow. The Master flow control limiter Hi-pots were set to provide a maximum flow (in Master Manual mode) of 985 as indicated on the B31-R613 recorder or 100% core flow per Process Computer Program CD-3 Option 2. The Level fd criteria associated with this test were met as follows:
- a. The computer code output indicated the following gain adjustment factors for B21-X602A/B jet pump flow loop sunmers at 91.82% core flow:
B21-K602A (GAF) = 0.9672 B21-X602B (GAF) = 0.9762 a Supplement 9 PQge 3.31-7 Composite GAFs for the jet pump loops were calculated per of Reactor Engineering Procedure 56.000.02 and were as follows: B21-K602A (CGAF) = 1.0~8 B21-K602B (CGAF) = 1.015 An evaluation run of 56.000.02 is planned at rated conditions to verify that this Level 2 criteria is indeed satisfied.
- b. The B31-K607 A,B,C,D summers (APRM/RDM flow-bias) were adjusted using the GAFs calculated in Reactor Engineering Procedure 56.000.02.
The previous GAFs were all set at 1.00 so the Composite Gain Adjustment Factors (CGAF) were equal to the calculated (GAFs). - The computei- code output indicated the following Gain Adjustment Factors for flow units B31-K607A ;hrough B31-K607D: Flow Unit A - B31-K607A (GAF) = 1.083 Plow Unit B - B31-K607B (GAF) = 1.092 Flow Unit C - B31-K607C (GAF) = 1.098 Flow Unit D - B31-K607D (GAF) = 1.077
- c. As previously noted, the mechanical stops for MG Sets scoop tubes were set at 86.6% speed (970 rps) for "A" and 85 7% (960 rpm) for "B". This translates to an equivalent flow of 102.5% for "A" and 102.2% for "B".
During Test Condition Six, another complete calibration of the recirculation system flow instrumentation was performed at 96.975 CTP and 96.54% CF to resolve issues raised in the previous core flow calibration. REP 56.000.02, Core Flow Calibration, was performed in the performance mode for this test. As a prerequisite, the zero and full span settings of single tapped and double tapped jet pump transmitters and square root converters were checked and calibrated as necessary. In addition, the Recirculation Flow Units were calibrated as required per REP 56.000.02 and the zera and full span settings of the Recirculation Flow Nozzle transmitters were obtained and documented. Supplement 9 Paga 3 31-8 A complete set of single tapped and double tapped l jet pump delta Pressure measurements were obtained, < and core flow evaluations were performed per this procedure. These evaluations revealed that the Loop Flow Variations for Loop A was 3 18% which exceeded ; the 35 criteria as stated. To resolve this issue, another set of measurements was obtained and evaluated per this procedure, and this time the Loop Flow Variation criteria for Loops A and B were 2.06% and 0.08%, respectively, which met the 35 criteria. This time, however, the Nozzle Plugging criteria, as stated in REP 56.000.02, associated with jet pumps 9/10 and jet pumps 11/12 was exceeded. To resolve this discrepancy, an attempt was made to obtain a I set of readings associated with jet pumps 9 and 10 i and l'or jet pumps 11 and 12 such that these readings l were uken at the same time every 10 seconds for 2 ' l minutes and averaged before they were compared. Based on this, the nozzle plugging criteria was satisfied. This is due to the fluctuating nature of these readings, and the outcome is quite sensitive , to how and when these readings are taken. De actual core flow was determined to be 96.54 Mlba/hr by performing the required calculations in REP 56.000.02. The indicated core flow on the control room recorder B21-R613 was reading 96 Mlba/nr and the process computer OD3 Option 2 edit was showing 97 73 Mlba/hr. ne loop drive flows at the time of this test were calculated to be 14.87 Mlbv./hr for Loop A and 15 06 Mlbm/hr for Loop B for a total drive flow of 29.93 Mlbm/hr. New M ratios were determined to be 2.159 for Loop A and 2.288 for Loop B. Total Rated drive flow was determined to be 31 Mlba/hr and the jet pump summers and flow units were adjusted based on the Gain Adjustment Factors derived in REP 56.000.02. When core flow was increased to 100% of rated core flow, the recirculation MG set speeds were measured using a Strobe-o-tac to determine how this data i point would compare with the MG Set speed versus core flow projection made during the pronous core flow calibration. The results were as follows: MG Set Speed (rps) A B 100% Core Flow 928 943 Existing Electrical Stop 953 948 I Existing Mechanical Stop 970 960 Supple: Int 9 P2ge 3 31-9 Using the projections obtained from previous test and superimposing the 100% core flow data point from this test on those points, the projected core flows associated with the electrical and mechanical stops are as follows: Projected Core Flows (%) MG Set A MG Set B Average At Zlectrical Stop 102.0 101.0 101.5 At Mechanical Stop 103 7 102.2 102 95 Based on this, both MG sets at their respective electrical stops would result in a projected core flow of 101.5% of rated. The MG sets at the mechanical stops would result in a projected core flow of 102.95% of rated. These are within the requirements of Technical Specification Surveillance 4.4.1.1.2 which states that the electrical and mechanical stops be set at less than or equal to 102.5% and 105.0% of rated core flow, respectively. Based on this, it was decided that the electrical and mechanical stops were not required to be adjusted. Also at 100% of Rated flow, the core dP was found to be 20.26 psid which is 1.61 psi below prediction which supports supplemental data analysis requirements. The Level 2 criteria associated with this test were met as follows: The computer code output indicated the following gain adjustment factors for B21-K602A/B at 96.54% core flow. B21-K602A (CAF) : 1.025 B21-K602B (GAF) 1.017 Com,xsite CAFs for the jet pump loops were calculated per Reactor Engineering Procedure 56.000.02 and were as follows: B21-K602A (CGAF) : 1.064 B21-K602B (CCAF) = 1.032 Jet Pump Suaners Gain Adjustments were made per REP 56.000.02 to satisfy this criteria. I Supplement 9 P:ge 3 31-10 The B31-K607 A,B,C,D sussers were cdjusted using the GAFs calculated in Reactor Engineering Procedure 56.000.02. The computer code output indicated the following Gain Adjustment Factors for flow units B31-K607A through B31-K6070: Flow Unit A B31-K607A GAF = 1.020 Flow Unit B B31-K607B GAF = 1.020 Flow Unit C B31-K607C GAF = 1.007 Flow Unit D B31-K607D GAF = 1.014 Composite GAFs for Flow Units B31-K607A through B31-K-607D were calculated N/ REP 56.000.02 and , were as follows: Flow Unit A B31-K607A GAF = 1.105 Flow Unit B B31-K607B GAF = 1.114 Flow Unit C B31-K607C GAF = 1.106 Flow Unit D B31-K607D GAF = 1.092 Adjustments were made per REP 56.000.02, and this Level 2 criteria has been satisfied. The mechanical stops for MG sets were set at 86.6% speed for "A" and 85.7% for "B". This translates to an equivalent flow of 103 7% for "A" and 102.2% for < "B" or an average of 102.95 for both which exceeds the criteria for this test but is within the requirements of Technical Specifications, and therefore no adjustments were made as previously discussed. This concludes the recirculation system flow l calibrations during the Startup Test Program. l l i l l I I Supplc :nt 9 Pago 3 32-1 3 32 Reactor llater Cleanup System 3 32.1 Purpose The purpose of this test is to demonstrate specific aspects of the mechanical operability of the reactor water cleanup system. 3 32.2 criteria Level 1 Nona Level 2 The temperature at the tube side outlet of the non-regenerative heat exchangers (NRHX) shall not exceed 1300 F in the blowdown mode and shall not exceed 120 F in the normal mode. The cooling water supplied to the non-regenerative heat exchangers shall be less than 6 percent above the flow corresponding to the heat exchangers capacity (as determined from the process diagram) and the existing temperature differential across the heat exchangers. The outlet temperature shall not exceed 180 F. The bottom head flow indicator will be recalibrated against the RWCU flow indicator if the deviation is greater than 25 gpm. The pump available NPSH is 13 feet or greater during the hot shutdown with loss of RPV recirculation pumps mode defined in the process diagrams. 3 32 3 Results During the Heatup test condition, the RWCU system was placed in a configuration so that flow was taken from the bottom drain and directly fed back to the < vessel, bypassing the demineralizers. In this configuration G33-610, bottom drain flow, should 1 read the same as G33-609, system inlet flow, our data showed a maximum deviation of 62 gpa. Bottom drain flow was recalibrated such that the Level 2 criteria could be satisfied. Also during Heatup, the RWCU syster was operated in both the normal and blowdown modes with the reactor at rated temperature and pressure. Process SupplenInt 9 Page 3 32-2 I variables were recorded in order to demonstrate the proper performance of the RWCU systen in each of these modes. The non-regenerative heat exchange tube side outlet temperatures for the normal and blowdown mode were 1120 F and 122 F tespectively. These values were within the Level 2 criteria limits of 120 F0and 130*F for each mode. Using temperature asasurements from the RBCCW side of the non-regenerative heat exchangers (NRHX) , the cooling water flow was calculated to be less ! than 6% above the NRHX capacity. The ; non-regenerative heat exchanger cooling water outlet l l temperatures were well within our Level 2 criteria ! of 180 F. Al". applicable Level 2 criteria were satisfied. ! During Test Condition Four, the Hot Standby Mode of the Reactor Water Cleanup System was used to prevent temperature stratification in the reactor vessel while the recirculation pumps were not running. Data was taken in this mode to determine the RWCU pump NPSH during natural circulation conditions ' where NPSH is most limiting. The actual NPSH was calculated to be 702 feet and is consistent with the > NPSH measured at other BWR plants. This satisfies the Level 2 criteria that NPSH be > 13 feet in the ~ most limiting mode of operation. This concludes the required Startup Test Program testing on the Reactor Water Cleanup System. I l , 1 f I I l i 2 _.___l Suppleeint 9 Pcge 3 33-1 3 33 Residual Heat Resoval System 3331 Purpose The purpose of this test is to demonstrate the ability of the Residual Heat Removal (RHR) Systen to remove residual and decay heat from the nuclear uys Ma so that refueling and nuclear servicing can be ptformed. 3332 Crlieria Level _1 wone Level 2 The RHE Fvstem is capable of operating in the suppression pool cooling and shutdown cooling modes at the flow rates and temperature differentials indicated on the process diagrams. 3 33 3 Results During the Heatup test phase, each division of the RHR systen was placed in the Suopression Pool Cooling Mode and process data was taken for a 30 minute time period. The extrapolated heat capacity for both heat exchangers indicated an excess capacity of 67 55. This was expected since in early heat exchanger life the heat transfer coefficient is larger and capacity was determined to accommodate some deterioration. The remaining testing in this section to demonstrate ther operation of RHR in the shutdown cooling mode has not yet been performed. It is anticipated that this test will be coordinated with the MSIV full isolation and Shutdown from Outside the Control Roon/ Cold Shutdown Demonstration tests. Supplement 9 PIge 3 34-1 3 34 Piping System Dynamic Response Testing 3 34.1 Purpose Verify that piping system structural behavior under probable transient loadings is acceptable and within the limit predicted by analytical investigations. 3 34.2 criteria Level 1 The measured vibration levels of the piping shall not exceed the acceptable specified values. Level 2 , The measured vibration levels of the piping must not exceed the expected specified values. 3 34 3 Results Piping dynamic transient vibrations were monitored during Heatup, in conjunction with Relief Valve , testing, for two SRV lines and selected Main Steam Lines. All vibration data recorded was within the , acceptable and expected limits as defined by the i Level 1 and Level 2 criteria. Piping dynamic transient vibrations were monitored during Test Condition Two in conjunction with relief valve actuations during relief valve testing, and during the planned Turbine / Generator Load Reject (Within Bypass) test. Data for the two SRV lines and the Main Steam Lines showed all vibration data was within Level 1 and Level 2 criteria except D-001, which was inocerable, and D-003 D-005 and D-008 which did not meet Level 2 criteria. All violations were reviewed and evaluated by Sargent and Lundy and were found to be acceptable. It is worth noting that the original criteria for these instruments were given as "information only" and were mistakenly incorporated into the procedure as Level 2 criteria. During Test Condition ihree, data was collected to determine the flow induced vibrational response of the High Pressure Coolant Injection (HPCI) system piping during a planned HFCI System cold vessel injection to the reactor. Supplement 9 Page 3 34-2 During the first successful HPCI cold vessel injection to the reactor the load on force pin F-155, located at the HPCI discharge, exceeded its Level 1 criteria. After a detailed walkdown of the HPCI supports and upon completion of further analysis of the HPCI Systen pipe supports by Sargent and Lundy, three additional strain gauge networks on three other HPCI supports were installed to monitor strains during the next cold injection. During that cold injection, all Level 1 and 2 criteria were satisfied. The additional strain gauges were monitored and these values were provided to Nuclear Engineering for evaluation and were found acceptable. Subsequent to this test, a HPCI vessel injection was performed on 7-5-87 which resulted in a HPCI overspeed trip. During that event, a water hammer and suction line overpressurization transient occurred (reference LER-87-030 00) which, after engineering analysis, has resulted in the replacement of E41-F005, HPCI Discharge Check Valve and several HPCI Systen hanger modifications. Due to these changes in HPCI piping configuration, this testing was reperformed to evaluate HPCI piping response during the next planned HPCI Quick Start testing sequence. That additional test was performed in conjunction with a Hot Quick Start Vessel Injection on October 14, 1987 and resulted in all criteria being met. The results from the three tests are summarized below Level 1 sensor 6/22/87 7/4/87 10/14/87 Allowable F-155 13966 lbf 4929 lbf 2164 lbf 10000 lbr As described in Reference 15 3 of this report, the initial test program was changed to allow taking credit for certain inadvertent scrans if a supporting analysis of the collected data demonstrated that the test results are valid when extrapolated to higher power levels. On December 31, 1987 an inadvertent turbine / generator trip did occur while at 74.8 percent reactor power. The UFSAR required analysis of the Cata collected from the turbine / generator Supples;nt 9 Pcg3 3 34-3 trip was subsequently performed by General Electric. The analysis demonstrated that although the trip occurred at a lower power level, the purpose of the test and all test criteria were satisfactorily met for the NSSS (GE) portion of the vibration data. This is documented in Reference 1.5.5 of this report. It should be noted also that the Sargent and Lundy required data was also recorded for this test. However, a failed sensor at the time of the inadvertent turbine trip induced noise in the sensor readings for the Sargent and Lundy data. Bis sensor, F-006, has been confirmed as the source of the unusual sensor readings and has subsequently been removed from service. Consequently, the Sargent and Lundy data recorded is suspect as are any associated criteria violations. The Sargent and Lundy vibration data will be recorded during a future turbine trip either inadvertent or scheduled. De Main Steam instrumentation in the drywell was unaffected by the failed F-006 and the data collected was found satisfactory by the General Electric NSSS piping analysis. During Test Condition Six prior to the entry into Test Condition Four, piping vibration data was recorded for the recirculation loops when Recire. Pump A was tripped with the reactor operating at 97% power. There were two Level 2 criteria violations that were evaluated by General Electric and found acceptable. The violations were as follows: Heasured Level 1 Level 2 Vibration Sensor Inches p-p ,1,nches n p-p Inches p-p RA-SX 1.626 0.038 0.202 RA-HY 0 318 c.008 0.029 After Recire. Pump B was reduced to minimum speed, vibration data was obtained when it was tripped to bring the plant to natural circulation conditions. All criteria was satisfied. Supplemant 9 Pcge 3 34-4 Subsequently, Reciro. Pump A was restarted on the ! recovery from Test Condition Four. Vibration data taken during the restart satisfied all criteria. The Recire. Pump B restart could not be performed at this time due to a torque switch problem with the Recirc. Pump B Discharge Valve, B31-F031B and will - be completed when the plant returns to power operation and this portion of testing is repeated. i j 4 i 1 l. k e I i l l ; 1
- s. Q@h sylvia Semor Vice Presedent
_. 6400 North One Higha af t-CISOn 37m";;'""'" Plant Technical Specifications September 20, 1988 NRC-88-0212 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555
References:
(1) Fermi 2 NRC Docket No. 50-341 Facility Operating License No. NPF-43 (2) Detroit Edison Letter to NRC "Startup Report Supplement 8" NRC-87-0136, dated June 20, 1988. . Subjects, Startup Report - Supplement 9 This is Supplement 9 of the Startup Report for Fermi 2. As required by Fermi 2 Technical Specification 6.9.1.3, a supplement is being submitted every 3 months until completion of the Startup Test Program. A supplemental report will be submitted by December 20, 1988. If you have any questions regarding this report, please contact Patricia Anthony, Compliance Engineer at (313) 586-1617. Sin ce r ely, C4 " cc: A. B. Davis R. C. Knop
- 7. R. Quay W. G. Rogers i i}}