ML20207B583
| ML20207B583 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde, 05000000, Trojan |
| Issue date: | 06/24/1988 |
| From: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Van Brunt E ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| References | |
| NUDOCS 8808020382 | |
| Download: ML20207B583 (132) | |
Text
- _ _
/
}o UNITED STATES g
g NUCLEAR REGULATORY COMMISSION i
j REGION V e,
1450 MARIA LANE,SulTE 210 WALNUT CRE E K.CALIFORNI A 94596 June 24, 1988 locket Nos. 50-528, 50-529 and 50-530 2
Arizona Nuclear Power Project P.O. Box 52034 Phoenix, Arizona 85072-2034 s,
Attention: Mr. E. E. Van Brunt, Jr.
Executive Vice President Gentlemen:
SUBJECT:
COMMENT CN ANPP AUXILIARY FEEDWATER SYSTEM DESIGN BASES MANUAL 4
By letter, dated March 10, 1988, you supplied for review a copy of the ANPP Design Base, Manual section on the Auxiliary Feedwater System and the associated sections from the Design Criteria Manual and System Descriptions Manual.
My staff and members of the NRR staff have reviewed these documents.
First, a few general comments are in order.
Your Detailed Design Criteria and System Description are heavily oriented toward the mechanical engineering aspects, and even so, do not provide any real detail regarding the rieans employed by the designers to implement, in the plant features, the very general design criteria.
In addition, we note that these contain iittle specific information regarding electrical and instrument.. tion and cont'.>ol system features implemented by the designers to effect tl.S general design criteria. These situations seem to make these documer.ts useful mainly as a reference to other documents, of limited usefulness to engineers performing modification design, and little usefulness to engineers engaged in monitoring r
and maintaining system performance in conformance to the intent of the desigr.er.
I am prov. ding for your information, by attacnment, a copy of the Portland General Electric Company Design Bases Document for the Trojan Nuclear Power Plant Auxiliary Feedwater System.
The scope and depth of this document were arrived at af ter much discussion between PGE and NRC. We feel this represents a creditable effort to capture the system design bases and provides a substantial reference for engineers engaged in designing system modificat%ns or monitoring and maintaining system performar.ce.
I wnuld urge sou to: critically compare your concept of a Design Bases Manual l
to the attached PGE concept, other indusi'y benchmarks, and the products of other utilities engaged in similar effort >; and come to an ANPP conclusion regarding the desired scope and depth of the ANPP Design Bases Manual.
.w6..,
'b OK 28 e
yo\\
e nu wMi&'""" $ '
- a. c v.u,
l RECElVED l
NRC
~2-REGION V I anticipate further discussion of this issue during one of our fut9re periodic management meetings.
88 JUL l, P i : 06 i
John B. Martin i
Regional Administrator Attachment-As stated 1
l cc:
J. G. Haynes, Vice President W. F. Quinn, Director, Nuclear Safety and Licensing i
R. Papworth, Director, Quality Assurance T. D. Shriver, Manager, Compliance bec w/er.21osures:
RSB/DocumentControlDesk(RIDS)
Resident inspector i
i 1
4 i
1 I
a i
i l
t i
j CONCURRENCES:
l N
f RPate/ow DKi ch i
rews C/. /88 6/g/88 6/pf/88
/g/88 I
/EQUESTCOPY, MUEST LOPY,
- UEST COPY,
ACQUEST COPY,
REQUEST COPY,
YE5 / NO YES / (MT)
YE5 / N0 YES / NO YES / NO l
b) f
{3DTOPDR 5
~D / NO i
i
--Ib6 [)
sh} fkfg
[
i j
certaried I;r_Tl*,u u c( de 0 u(C>-
m I
i i
9 G-C) s l-I
O O
TROJAN NUCLEAR PLANT AUXILIARY FEEDWATER SYSTEM DESIGN BASIS DOCUMENT CONTENTS
_Section Title M
1.0 FUNCTIONAL DESCRIPTION.
1-1 1.1 Purpose 1-1 1.2 System Boundaries 1-1 1.2.1 Mechanical....................
1-2 1.1.2 Electrical....................
1-3 1.2.3 Instrument and Control......
1-4 1.2.4 System Interfaces 1-5 1.2.4.1 Diesel Fuel Oil System..
1-5 1.2.4.2 Condensate Makeup Water System.
1-5 1.2.4.3 Service Water System.
1-5 1.2.4.4 Feedwater and Condensate System 1-6 1.2.4.5 Main Steam System 1-6 1.2.4.6 Reactor Protectior. System 1-6 1.3 System Performance Recuirements 1-7 1.3.1 American Nuclear Society Conditions 1-7 1.3.1.1 Loss of Normal Feedwater..........
1-8 1.3.1.2 Major Rupture of a Main Feedwater Pipe.
1-10 1.3.1.3 Induced Malfunction in the Steam Generator PORV Control System.
1-10 1.3.1.4 Induced Malfunction in the Pressurizer PORY Control System.
1-11 1.3.2 other Performanci Requirements.
1-13 1.3.2.1 Control Room Inaccessibility.
1-13 1.3.2.2 Loss-of-Coolant Accident Concurrent with Leaking Steam Generator Tube.
1-13 1.3.2.3 System Cooldown 1-13 1.3.3 Mode operability.
1-16 1.3.4 Summary 1-16
[
2.0 CODES, STANDARDS, AND REGULATORY DOCUMENTO.
2-1 2.1 Ceiteral Design criteria 2-1 I
2.2 Regulatory Documenta.tlen.............
2-2 l
I 2.2.1 Code of Federal Regulations (CTR) 2-4 2.2.2 Regulatory Guides 2-3 2.2.3 KURECs....
2-4 E,
. :. a.w a.
- w d
e TROJAN NUCLEAR PLANT AUXILIARY FEEDWATER SYSTEM DESIGN BASIS DOCUMENT CorTENTS
_8)3 tion.
Title Eggt 2.3 godes and Standards 2-4 2.3.1 General 2-5 2.3.2 Mechanical....................
2-5 2.3.3 Electrical....................
2-6 3.0 SYSTEM DESICW BASES 3-1 3.1
'Finina Pressure and Temperature Criteria.
3-1 3.2 Seismis criticia..
3-1 3.2.1 P-187.
3-3 3.2.2 Recirculation Lines 3-3 3.2.3 CST 3-3 3.3 Quality Assurance criteria..
3-4 3.4 Redundancy / Diversity Criteria 3-5 3.5 Environmental cualificatlan Criteria.......
3-8 3.6 T.tre Protection Criteria...
3-11 3.7 Environmental Protection Criteria 3-11 3.).1 Flooding Protection 3-12 3.7.1.1 External Flooding 3-12 3.7.1.2
' Internal Flooding 3-13 3.7.2 Missii9 reotection.........
3-13 i
3.7.3 To rnado Pco t e c t ion................
3-14 l
3.7.4 Pipe Whip and Jet Imp!.ngement 3-15 q
3.8 Automatic Initiation........
3-36 4.0 C0KPOWENT DESIGN BASES.
4-1 4.1 ArW Pumps P-102A and 4-1023 4-1 4.1.1 Design Requirements 4-1
)
4.1.2 Configuration
.-2 4.1.3 Margin Evaluation 4-3 c,.
'.. n.;.ia.
' r, s G w,
2.:
TROJAN NUCLEAR PLANT AUXILIARY FEgDWATER SYSTEM DESICW BASIS DOCUMENT CONTENTS Section Title E111 4.2 Turbine Driver K-107A 4-3 4.2.1 Design Requirements 4-3 4.2.2 Configuration 4-4 4.2.3 Martin Evaluation 4-7 4.3 Diesel Driver K-1075.
4-7 4.3.1 Design Rnquirements 4-7 4.3.2 Configuration 4-8 4.3.3 Margin Evaluation 4-10 4.4 AFW Pump P-182.
4-10 4.4.1 Design Requirements 4-10 4.4.2 Configuration 4-11 4.5 Manual valves 4-12 4.5.1 Design Requirements 4-13 4.5.2 Configuration 4-13 4.6
' Auxiliary Feedwater Flow Control Valves 4-11 4.4.1 Design Requirements 4-14 4.6.2 Configuration 4-14 4.7 Turbine Steam Supply isolation Valves 4-15 4.7.1 Design Requirements 4-15 4.7.1.1 CV-1451 through CV-1454 4-15 4.7.1.2 NO 3170 4-16 4.7.2 Configuration 4-17 4.7.2.1 CV-1451 through CV-1454 4-17 4.7.2.2 MO-3170 4-18 s
4.8 Turbine Trip and Throttle Valve 4-18 4.4.1 Design Requirements 4-18 4.4.2 Configuration
-19 4.9 Servlee Water System Isolation Valves 4-20 4.9.1 Design Requirements 4-20 4.9.1.1 MO-3045A and MO-3045B 4-20
- 111 -
., ;s...
. L a w wi....
pc :r).
n
O TROJAN NUCLEAR PLANT AUXILIARY FEEDWATER SYSTEM DgSICW BASIS DOCUMENT CONTgNTS section Title tagg 4.9.1.2 MO-3060A and MO-3060B 4-21 4.9.2 Configuration 4-21 4.9.2.1 ho-3045A and MO-30458 4-21 4.9.2.2 MO-3060A and H0-30608 4-22 4.10 Condensate storate Tank Level Inctrumentation 4-23 4.10.1 Design Requirements 4-23 4.10.2 Configuration 4-24 4.11 Differentist Pressure control 4-25 4.11.1 Design Requirements 4-25 4.11.2 Configuration 4-26 4.12 Flow Indication 4-27 4.12.1 Des,ign Requirements 4-27 4.12.2 Configuration 4-27 4.13 Remote Shutdown station 4-28 4.13.1 Design Requirements 4-28 4.13 2 Configuration 4-28 4.14 AFW Pump P-182 Discherme_ Isolation valves 4-29 4.14.1 Design RequirementE 4-29 4.14.2 Configuration 4-29 4.15 ATW Pump P-182 Differential Pressure Control Valve 4-30 4.15.1 Design Requirements 4-30 4.15.2 Configuration 4-30 5.0 SYSTEM OPERATION.
5-1 5.1 Wormal operations 5-1 5.2 Woreal T*ansient operations 5-1 5.3 Abnormal and Emergency operations 5-2
- tv -
.. w
.c: A g,
TROJAN NUCLEAR PLANT AUEILIARY FEEDWATER SYSTEM DESICM BASIS DOCUMENT CONTENTS 1
Section Title
- EAL9, 6.0 INSPECTION AND TESTING.
6-1 6.1 Deslan Inspection and Testing Requirements.
6-1 i
6.2 31 stem Inspection and Testint 6-2 7.0 DESIGN BASES EVOLUTION.
7-1 7.1 Completed and Closed Out RDCs 7-1 j
7.2 RDCs Installed But Not Closed out 7-10 i
7.3 RDCs Incomplete or Pendinz............
7-11 i
8.0 FIGURES 8-1 9.0 TABLES.
9-1 i
10.0 REFERENCES
10-1 10.1 Codes. Standards. and Regulatory References 10-1 i
10.1.1 Codes and Standards 10-1 10.1.2 Regulatory Refereness 10 2
)
10.2 PCE Internal Doctanents and Technical Manuals...
10-6 10.2.1 PCE/Trojon Reports................
10-6
- 0.2.2 Technical Mar References 10-9 10.2.3 Trojan Plant.';trating Manual (POM References) 10-9 l
10.3 PCE/ Trojan Prints. Drawltirs. and Specifications 10-10 10.3.1 Specifications.
10-10 i
i 10.3.2 Piping and Instrumentation Diagrams (P& ids) 10-10
[
10.3.3 Logic Diagrams.
10-10 t
l 10.3.4 Electrical Prints 10-11 10.4 Vendor Reports. Specifications and Miscellaneous References....................
10-13 i
t I
~
l 1
i
-V-
- 1~
c'
'rk' slA
s**%v
's q f.s
.x s-w
TROJAN NUCLEAR PLANT AUXILIARY FEEDWATER SYSTEM DESIGN BASIS DOCUMENT COWTENTS i
- ection Title P,,a_q, 10.5 Correspondence References 10-14 10.5.1
. Vendor Corsespondence 10-14 10.5.2 Regulatory Correspondence 10-15 l
10.5.3 Internal Correspondence. '.
10-17 10.6 Calculation References...
10-18 10.6.1 PCE Calculations..
10-18 10.6.2 Bechtel Calculations..........
10-20 Appendix 1 OPEN ITEMS.
A-1 t
l i
i I
9 h
I i
l l
- vi-
~
N
J# *I Id
t 7
j I
r TROJAN NUCLEAR PLANT l
4 t
AUXILIARY FERDWATER SYSTEN i
DESIGN BASIS DOCUMENT l
1 j
CONTENTS i
2 9
L List of Finures j
8-1 Auxiliary Yeedwater System i
j t
l
[
1 r
t I
t t
[
l l
i t
'a i
t h
i 1
i I
I i
f f
i t
f l
y
- vii -
I
- .. a...
..., ;,i
.+
e TROJAN NUCLEAR PLANT AUKILIARY FEEDWATER SYSTEM DESIGN BASIS DOCUMENT e
CONTENTS List of Tables 9-1 AFWPumpsh-102AandP-1028DataSheet 9-2 Turbine Driver K-107A Data Sheet 9-3 Diesel Driver K-1075 Data Sheet 9-4 AFW Pump P-182 Data Sheet 9-5 AFW System Instrumentation at the Remote Shutdown Station (C-160) e 9
e
- viii -
- :.... -4.
.. :. % :w
TROJAN NUCLEAR POWER PLANT AUIILIARY FEEDWATER SYSTEM DESIGN BASES DOCUMENT 1.0 FUNCTIONAL DESCRIPTION l
121 pURp03E The safety function of the Auxiliary Feedwater (AFW) System is to supply waterto the secondary side of the steam generators for reactor decay heat removal when normal feedwater sources are unavailable.
The AFW System also supplies feedwater to the steam generators during startup, hot standby, and shutdown.
1.2 SYSTEM BOUNDARIES l
The ATW System boundary from suction to discharge (including the water j
source and heat sink) incli:Jes those portions of the system required to accomplish the ATW system function and connected branch piping up to and including the second valve, which is normally closed or capable of automatic closure when the safety function is required. The AFW System boundary also includes (as described in a Nuclear Regulatory Commission (NRCl Safety Evaluation Report (SER) dated August 4, 1981) any portion of branch piping that is structurally coupled to the ATW System boundary in a way that causes the seismic response of the branch piping to transmit loads to the AFW System. As a minimum, this includes the branch lines outside the AFW System boundary to a point l
of three orthogonal restraints. All mechanical and electrical equipment, piping (eg, instrument air), conduits, and cable trays that are necessary for, or contain items that are necessary for, the operation of the AFW System are also considered within the bounds of
{
the ATW System.
In addition, the structures housing these systems and I
components are included.
Similar constraints apply to alternative i
1-1
. a.. u i
-u
. s..c.
?
means of decay heat removal. The mechanical, electrical, and instrument.and control components that are considered a part of the AFW System boundaries are listed in the following subsections.
d Figure 8-1 illustrates the AFW System boundaries.
t 1.2.1 MECHANICAL i
(1) AFW Pumps P-102A and 1028.
(2) DeLaval oli pump.
(3) Tdebine Driver K-107A.
(4)
Turbine Driver V.-107A drain coolers.
(5) Turbine Driver K-107A and P-102A lube oil cooler.
(6) Diesel Driver K-107B.
l (7) Diesel Driver K-1078 speed increaser I-373.
(8) AFW pump diesel fuel oil day tank (T-152).
s (9)
Pump P-1028 coolers.
(a) Diesel lube oil cooler.
(b) Jacket water cooler.
I (c) Speed increaser gear lube oil cooler.
(d)
Diesel intercooler.
(e)
Pump P-1020 lube oil cooler.
(10) Electric AFV Pump P-192.
i (11)
Lube Oil Pump P-183.
1-2 L _,...g.t 4;.
1 (12) Pump P-182 shaft-driven oil pump.
(13) Manual valves.
(14) AFW flow cor: trol isolation valves (CV-3004 Al through D1 I
and A2 through D2).
(15) Turbine steam surply isolation valvas (CV-1451 through CV-1454 and MO-3170).
(16)
Solenoid Valves SV-1451 through SV-1454.
(17) Accumulators T-166A through T-166D.
(18)
Turbine Trip and Throttle Valve MO-3071.
(19) service Water System Isolation Valves (MO-3045A, MO-3045B.
MO-3060A, and M0-30608).
(20) P-182 motor-operated discharge isoletion valves (MO-2947A and NO-29472).
(21) P-182 differential pressure control valve (CV-2967).
1.2.2 ELECTRICAL (1) Remote shutdown panel (C-160).
(2) Turbine Driver K-107A governor.
(3) Diosal Driver K-1078 starter battery and battery charger.
(4) Diesel privo? K-1075 governor.
1-3 s u:. :
(5) pump P-182 ac motor.
(6) Valve motor operators.
1.2.3 INSTRUMENT AND CONTROL (1) Limit switches for valve position indication in control room' and Panel C-160.
(2) Flow Switches FIS-3004 Al through D1 and FIS-3004 A2 through D2 and automatic flow control valve closure and lockout controls.
(3) Condensate storage tank level instrumentation.
i (4) ArW pump differential pressure conirot circuits.
(5) Valve position switches.
(6) Turbine Driver K-107A steam inlet and exhaust pressure
. monitoring loops.
(7) Pump P-102A suction and discharge pressure monitoring loops.
(8) Pump P-1028 suction and discharge pressure ac91toring loops.
(9) Diesel Fuel 011 Tank T-152 level controls.
(10) AFW flow monitoring instrumentation.
1 (11) Turbine Driver K-107A speed monitor loop and electrical overspeed trip.
)
~
1 1-4 s
4
(12)
Pumps P-102A and P-102B, and Speed Increaser I-173 bearing temperature monitoring loops.
(13)
Turbine Driver K-107A local temperature and pressure sages.
(14) P-102A local temperature and pressure gages.
(15)
Diesel Driver K-1075 local temperature, preosure, and speed sages, ar.d local temperature and pressure sages.
1.2.4 SYSTFM INTERFACES The following paragraphs describe the systems that interface with the ATW System.
1.2.4.1 Diesel Fuel Oil System The Diesel Fuel Oil System supplies fuel oli to the diesel-driven AFW s
pump fuel oil day tank (T-152).
't 1.2.4.2 Condensate Makeup Water System The SCII condensate storage tank (CST) is the preferred source of water for the diesel-driven and turbine-driven AFW pumps.
It is the L
I only source of water for the electric AFW pump. Recirculation flow from each AFW pump is directed back to the CST.
1.2.4.3 Service Water, System I
t i
The Service Water System is the emergency water supply to the suction of the diesel-driven and turbine-driven AFW pumps.
The Service Water System also supplies cooling water to the diesel-driven AFW Pump.
service water cocls the diesel jacket coolert the diesel engine intercoolert the diesel engine and pump lube oil coolers and the gear t
1-5
+
s
e Lube oil cooler on the speed increaser.
Service water is used as the
'oackup coolant for the turbina-driven AFW pump lube oil cooler and the turbine bearing heat exchangers.
1.2.4.4 Feedwster and t'ondensate Systeg The AFW System supplies, emergency feedwater to the steam generators through Feedwater System piping at the inlet to the steam generators.
The electric AFW pump supplies feedwater through the Feedwater System piping to the steam generators during normal plant startup and shutdown.
If. both main feedwater pumps trip, a signal automatically starts the ESF AFW pumps to supply feedwater to the steam generators.
The Condensate Chetnical Injection System inlet isolation valves and pumps are manually opened and are started at Chemistry Department request to supply hydrazine to main feedwater lines entering the Containment when using the electric AFW pump to place the steam generators in a wet layup condition.
1.2.4.5 Main Steam System Steam is supplied to the Terry turbine of the turbine-driven AFW pump from the Main Steam System.
Steam supply 13.nes tap off each main steam header upstream of the main steam isolation valves.
During ATW System operation. steam pressure signals from the main steam headers are compared to the discharge pressure signals from the AFW pumps to control ATW flow to the steam generators.
1.2,4,6 Res; tor protection System The Reactor Protection System provides automatic starting signals to th.s diesel-driven and turbine-driven AFW pumps.
These signals 9
1-6 e
Nt
(Channel A for the turbine-driven pump and Channel B for the diesel-driven pump) are (1)
Safety injection (SI) (any SIS).
l (2)
Steam generator low-low level (il percent, 2/3 any steam gene ra t o r)..,
(3) Undervoltage on 4.16-kV Suses A1/A2 (undervoltage on degraded grid).
b 1.3 SYSTEM PERFORMANCE REQUIREMENTS This subsection describes the design, protective, and operational outputs of the AFW System as a function of operating mode or condition.
Included is a description of the physical conditions under which such operation is required to occur.
1,3,1 AMERICAN NUCLEAR SOCIETt CONDITIONS i
Since 1970. Westinghouse has used the American Nuclear Society (ANS) i classification of plant conditions, which divides plant conditloas into four categories according to anticipated frequency of occurrence and potential radiological consequences to the public.
The four t
categories aret L
(1) Condition It Normal operation and operational transients.
(2) Condition II:
Faults of moderate frequency.
5 l
(3) Condition III:
Infrequent faults.
t (4) Condition IV:
Limiting faults.
(
1-7 1
I
.. l: e
The AFW Tystem is designed to maintain its functional capabillsy when any of the following conditions occurst (1)
Loss of normal feedwater.
(2) Major rupture of a main feedwater pipe.
(3)
Induced malfunction in the Steam Generator PORY Control Cystan.
(4)
Induced malfunction in the Pressurizer PORY Control System.
1.3.1.1 Loss of Normal Feedwater A loss of unreal feedwater (from pump or turbine failures, valve nalfunctions. or loss of off-site ac power) is an ANS Condition II event.
If an alternative supply of feeddater were not supplied to the Plant daring this accident, residual heat after the reactor trip would heat the primary system water enough to cause water relief from the pressurizer.
Significant loss cf coolant from the Reactor Coolant System (RCS) could lead to core damage. Analysis shows that after a loss of all off-site ac power simultaneous with a loss of normal feedwater, the Atw System can remove the stored and residual heat, and thus prevent overpressurization of the RCS ano loss of coolant from the reactor core.
The BLK0UT code (see WCAP-)t98. Long Term Transient Analysis for PWRs) was used in the analysis to model the plant trtnsient after a loss of normal f6 iwater.
Major assumptions were (1)
The initial steam generator water leve (in all steam generators) at the time of reactor trip is at a conse.vatively low level, ie. the lower narrow-range level tap.
1-8 A'-
(2) The Plant is initially operating at 102 percent of the engineered safeguards design rating.
(3) A conservative core res* dual heet generation is based upon long-term operation at the initial power level preceding the trip.
(4) A heat transfer coefficient in the steam generator is associated with the RCS natural circulation.
(5)
Auxiliary feedwater flow is available 1 min. after the accident, at 426 spm.
(6) Auxiliary feedwater is delivered to two steam generators.
(7)
Secondary system steam re?lef is achieved through the code safety valves.
Steam relief is throush the power operated relief valves'(PORVs) or steam dump valves for most cases of loss of notnal feedwater.
However, analysis assumed these to be unasalla' ole, i
(8)
The initial reactor coolant average tamperature is 4'F lover than the nominal value since this condition results in a greater ervansien of RCS coolant during the transient and in a higher water level in the pressurizer.
The results of this analysis are presented in Figure 15.2-9 of the UTSAR which shows plant parameters af ter a loss of normal feedwater.
At no time is the tube shewt uncovered in the st.eam generators receiving AFW flow, and at no time le there wr.ter relief from the pressurizer.
Thus, if the auxiliary feed delivered is greater than that assumed (426 spm), the initial reactor power is less than 102 percent, or if the steam generator water level in at least one i
steam generator is above the low-low level trip point at the tine of i
1-9 J
i i
the trip, the result will be a steam generator minimum water level l
higher than shown and increased margin to the point at which reactor coolant water relief occurs.
[
i t
_1.3.1.2 Maior Ruoture of a Main Feedwater Pigg A major feedwater line, rupture (ANS Condition IV event) is defined as I
a break in a feedwater pipe large enough to prevent the addition of E
sufficient feeduater to the steam tenerators to maintain shell side i
fluid inventory in the steam genetators.
If the break is inside containment.between the check valve and the steam generator, fluid j
from the stets generator may also be discharged through the break.
[
Further, a break in this location could preclude the subsequent i
addition of AFW to the affected steam generator.
Analysis (see WCAP-7909 MARVIL-A Digital Computer Code for Transient Analysis for a Multiloop PWR Systems showed that for the postulated feedline rupture, the assumed tsFW System capacity is adequate to remove decay heat, prevent overpressurizing the RCS, and prevent uncovering of the reactor core.
The ArW is assumed to be initiated 10, min. after the trip with a feed rate of 426 gym. An additional 5 min. is assumed to elapse before the feed lines are purged and the reintively cold (120*F) AFW enters the unaffected steam generators.
The assumed AFW flow rate can remove decay heat 2.100 seconds after a trip. After this time, core decay heat decreases below the AFW heat removal capacity, and reactor coolant temperature and pressure decreasew.
1.3.1.3 Induced Malfunetton in the Steam Generator PORY control System Af ter a feedline rupture outside Containment, the steam generator p0RVs are assumed to exhibit a consequential failure due to an adverse environment.
Failure of the PORY in the open position results in the 1-10
0 0
i depressurlastion of. multiple stema generators, which are the source of steam for the steam turbine-driven ArW pump.
This scenario was analysed (see UFSAR Section 15.2.3.3) with the following assumptions:
f t
(1) A break occurs outside Containment betseen the penetration and feedline check valve.
i (2) An adverse environment resulting from the rupture affects the steam generator PORV control systems associated with 6
the ruptured loop and the intact loops.
L (3) A single active failure occurs in the diesel engine-driven i
ArW pump.
Analysis of this scenario demonstrated that postulated break locations at Trojan would be limited to 5-to 10-ft lengths of 14-in.-diameter feedwater pipe cdjacent to the containment penetrations.
These piping runs are located entirely within the compartmentalized Main Steam Support Structure (MSSS). A main feedwater piping rupture of the class that leads to this scenario could not generate an adverse environment at the location of the steam generator PORY components 1
associated with any of the intact loops. At most, the PORY control system for the affected steam generator could be subjected to an i
adverse environment.
In this case, even if the affected PORY were to
{
fall open, steam delivery to the AFW pump turbine driver from the i
three unaffected steam generators would continue to be available.
[
1.3.1.4 Induced Malfunction in the pressurizer PORY Control System l
As part of the follow-up efforts to the Three Mile Island 2 (TMI-2) b accident Westinghouse analyzed this class of accidents and repcrted t
the results in WCAP-9600 Report on Small Break Accidents for, Westinghouse NSSS Systems. The analyses assumed a total loss of I
feedwater with various concurrent small primary Pipe breaks.
The i
1-11
6 transients were analyzed to 5000 seconds, without operator action end with an assumption of no auxiliary feedwater, to determine when operator action would be required to ensure no core uncovery.
The WCAP-9600 analyses concluded that the worst-case situation would be an optimally sized break that just precluded delivery of safety injection fluid to the RC3.
This break size (about 0.2 in. in diameter) is considerably smaller than the open area of one of the two i
pressuriser PORVs. Section 4.2.3.5 of WCAP-9600 indicates that no i
operator action would be needed if both pressurizer PORVs were to fail open since safety injection flow would then be sufficient to ensure no core uncovery.
Therefore, the conclusions reported in WCAP-9600 are conservative with respect to the possibility of one or both pressurizer PORVs being stuck open because of consequential malfunction of the PORY control system.
Furthermore, the WCAP-9600 analyses were based on a typical two-loop plant.
This basis ensures a conseevative minimum action time because of the small RCS water inventory relative to a four-loop plant such as Trojan.
The conclusion reached in Section t.2.4 of WCAP-9600 is that for the worst-case primary pipe break (0.2-in. equivalent diameter) concurrent with the loss of all feedwater, the plant can be brought to a fully stable situation without core damage, provided AFW flow is initiated within 3.500 seconds.
l To apply these conclusions to the case of a concurrent foodline rupture. Westinghouse performed additional calculations f
conservatively, and assumed that all liquid inventory in the steam generator associated with the ruptured feed line would flow out of the rupture without removing any heat (ie, liquid blowdown). These calculations show that the heat removal capability of the liquid l
inventory blowdown requires operator action 1.200 seconds earlier than the time reported in WCAP-9600. Thus. if a feed line rupture is assumed to be coincident with the analyses performed in Section 4.2 of WCAP-9600, the operator has at least 2.300 seconds to cause injection 1-12
O e
cf AFW into the intact steam generators. UFSAR Section 15.2.8.1 assumes ATW initiation at 60 secondst hence, the consequences of feed line rupture with the consequential failure of the pressurizer PORV control system are bounded by those reported in Section 15.2 of the UFSAR.
1,3,2 OTHER PERFORMANCE REQUIREMENTS 1.3.2.1 Control Room Inaccessibility The remote shutdown station for AFW System operation is required to assure the Plant can be brought to a safe condition after a main control room evacuation (Westinghouse to Bechtel Letter POR-1472 October 20. 1972).
Sufficient controls and indications are available at the Remote Shutdown Panel (C-160) to initiate and monitor AFW System performance.
1,3,2,2 Loss-of-coolant Accident Concurrent with Leakinr Steam Generator Tube In this case, a barrier is maintained between the steam feed lines through preservation of a pressure differential that prevents leakage from the RCS into the secondary plant.
Immediately after the loss-of-coolant accident, the steam pressure is above the RCS pressure.
When the RCS pressure is low, the differential pressure is maintained by a static head above the leaking tube.
The AFW Syrtem provides water to fill the steam generator and thus prevents leakage from the primary to secondary side.
1,3,2,3 System Cooldown The AFW system must supply sufficient feedwater to the steam generators to remove decay heat and maintain hot standby conditions for 2 hr end subsequently cool down the primary system to the 1-13
A
,/4O g
+
e, A. -
o rs
,f,,%
O t
IMAGE EVALUATION
\\/////
g' I/
[g9/7 ch t,
TEST TARGET (MT-3)
\\f y
,qq, i
i 1.0 20 ill- =
- 1. 8, IL s!.}.k !i l.6h = c ::
4 150mm 4
~
6' j
4 N
b.-
h.
y
<[E \\
Q
~
- g h,*.;k%
<x.
g)7,7,y
s4
'o 7;
A__
t$
e&
O v@V. 49 e
v 9
O tWN IMAGE EVALUATION
- (# ~
43 N'///// **i *** %
//fq,
- 4$*
'"("'~)
9
\\% //
7 $$1
////
p l.0 E L% 2 5!NO 1.1 ( E l
M l.8 bl=:s 1.2 5 i
1 ir 1.6 I a m.4L 4_._
150mm l
l 4-6" i
m, 4 $,
byp q&
- 4 zzzz m
n,e g zs gp i
s 3
t, u, o
.O
'e 4%
[
Qt"
$('t* '
f
. j,
~
t I
O-)
{
.& *e 0
(
% V 'o%
- e..
//j#
[<.its,O O
t&V IMAGE EVALUATION
\\///// 1[
$2rI9 TEST TARGET (MT-3)
/
}/
l.0 b E BA y tu pyy c 1::,
l,l h
bO I
iH 1.8
=
w:1t3 1.25 l I.4 l' L6
_a Ir m i!m 150mm 4
6" Y
,,[ g _g
,, q o r /-A
/>
O
/..
4 9 o^
5
.;f@p s
Of sp '
~
ed.
o l
temperature and pressure at which the Residual Heat Removal (RHR)
Sys'em can operate (a cooldown to 350*F, which corresponds to a steam generator pressure of 125 paia) within 4 hr.
The design basis AFW flow rate to support this cooldown la 880 gym.
Westinghouse design criteria assume a minimum AFW flow rate of 440 gpm to envelop the accident analysis assumption of 426 spa available flow.
The minimum rated flow per pump required by Westinghouse to maintain sufficiesit heat transfer surface in the steam generators while limiting reactor coolant temperature rise to prevent relief of water through the pressurizer relief valve is 440 gym (per Westinghouse to Bechtel Letter p0R-121 dated June 4, 1969).
- However, a 440-spm capacity would lengthen the time for reactor cooldown.
l Hence, Westinghouse recommended a tet:1 =inimum available AFW flow rate of 880 spm for the hot shutdown and turbine trip conditions to meet the 4-hr cooldown criteria (per Westinghouse to Bechtel Letter POR-113 dated June 3, 1969).
The consequent system design specifies a minimum flow rr's of 880 gpm per pump in order to envelop j
both this requirement and oth.e single failure considerations.
TheJminimum usable condensate storage supply of feedwater required by original Westinghouse design (per Wastinghouse to Bechtel
]
Letter POR-728 dated Harch 29, 1971) to effect this cooldown was 100,000 gallons. The Westinghouse calculations supporting this volume are noi, available to pCE. However, the assumptions used to determine this voluma were:
(1)
Cooldown rate of 50*F per hour.
(2)
Reactor coolar.t pumps shut down.
(3)
No water being ysed exc2pt to feed the steam generators.
1 (4)
No blowdown from the stean generators.
1-14
(5) No reserve tank volume available.
(6) Storage requirements for feedwater makeup or initial charging of the system not included.
(7) Plant to be cooled down from maximum calculated load to hot standby, held at hot standby for 2 hr, then cooled down to 350*F in 4 hr.
1 Reanalysis of the required condensate storage volume was provided by Westinghouse in 1973.
This volume of 196,000 gallons is illustrated
}
by Westinghouse Curve SSE-111t dated Decer.ber 11, 1972. Additional assumptions used'in determining Curve SSE-1119 were:
(1)
Steam generator levels are initially at the low-low level, i
(2)
Steam generators are refilled to no-load prograrrened level at completion of the cooldown to 350'F.
(3) Cond4nsate water temperature is 100*F.
4 s
l The design basis usable volume of 196.000 gallone is greater than that l
determined previously due to increased margin required by Westinghouse l
to account for uncertainty in plant metal mass (see Westinghouse to Bec' c 1 Letter POR-1613 dated Febniary 14, 1973).
Westinghouse evaluation of this requirement for Trojan (per Westinghouse to Bechtel Letter POR-85-579 dated May 24, 1985) assumes delivery of the required volume to the active steam generators and includes no allowances for unusable CST volume. AW pump NPSH requirements. or any condensate water lost fro: a pipe break prior 'o l
1 solation.
i l
I I
1-15
1.3.3 MODE OPERABILITY OPERABILITY of the AFW System ensures that the RCS can be cooled to less than 350*F from the normal operating temperature in the event of a total loss of offsite power. Either the diesel-driven or the turbine-driven AF'd pump has the capacity to provide sufficient feedwater flow to remove reactor decay heat and reduce the RCS pressure and temperature to 400 psig and 350*F. respectively, at which point the Residual Heat Removal System may be placed into operation for continued cooldown.
Limiting Condition for 0poration I
(LCO) 3.7.1,2 requires that at least two independent AW pumps and associated flow paths be OPERABLE in MODES 1, 2, and 3 with:
(1) One feedwater pump capable of being powered by an OPERABLE J
diesel with g450 gal. of fuel in its day tank.
(2) One feedwater pump capable of being powered from an OPERABLE steam supply system.
t 1
)
The AW System is not required to be OPERABLE in MODES 4
- 5. or 6.
I i
4 t
1.3.4
SUMMARY
The AW System is designed to supply feedwater at a flow sufficient to ensure adequate heat transfer area coverage in the steam generators to j
prevent a temperature rise in the reactor coolant that would cause release of coolant through the pressurizer relief valves.
This condition is satisfied as long as the AW System can provide feedwater at a minimum flow rate of 426 gpm total to two steam generators within 1 min. after receipt of a signal that automatically starts the A W pumpd.
l AW System flow is required to remove decay heat and cool dodm the RCS to 350'F corresponding to a steam generator pressure of 125 psia, at which time the Residual Heat Removal System can be operated.
The 1-16
J e
system must also be able to provide the required AFW flow for at least 2 he from~one AFW pump train, independent of any ac power source (per GS-5 of NUREG-0611).
In addition, suf ficient f sedwater (196.000 gallons of usable water in the CST) must be available to maintain hot standby conditions for 2 he after a reactor trip, then cool down the primary system at an average rate of 50'T per hour to the temperature and pressure at which the Residual Heat Removal System
[
can operate (a cooldown to 350*F in 4 hr).
L F
k r
s l
l
{
i l
t r
i 1-17
2.0 CODCS, STANDARDS, AND REGULATORY DOCUMENTS This section describes the regulatory documents, codes and standards, and general design criteria (CDC) applicable to the design, procurement, manufacture, installation, testing, operation, modification, and maintenance of the Auxiliary Feedwater (AFW) System and its components. With the exception of the CDC listed in 10 CFR 50 Appendix A and described in Updated Final Safety Analysis Report (UFSAR) Chapter 3. Section 3.1, regulatory documents, codes, and standards cite the specific component or portion of the AFW System to which they apply.
2.1 CEKERAL DESIGN CRITERIA The (c11owing CDCs are applicable to the AFW system and its components, electrical supplies, and instrumentation (1) Criterion 1:
Quality Standards and Records.
(2)~ Criterion 2:
Design Bases for protection Against Natural phenomena.
(3) Criterion 3:
Fire protection.
(4) Criterion 4:
Environmental and Missile Design Bases.
(5) Criterion 13:
Instrumentation and control.
(6)
Criterion 17:
Electric power Systems.
(7) Criterion 19:
Control Room.
2-1
t e
a i
\\
{
{
(8) Criterion 20:
Protection System Functions.
J k
(9) Critorion 34:
2.2 REGULATORY DOCUMENTATION e,
The following regulatory documents are applicable to the Auxiliary i
i Feedwater (AFW) Sy8 tam or its components, electrical power supplies, and instrumentation as noted.
t l
i 1
2,2.1 CODE OF FEDERAL REGULATIONS (CFR1 i
s (1) 10 CFR 50.48.
Fire Protection (AFW System and componants).
1 l
(2) 10 CFR 50.49.
Environmental Qualification of Electric.
I 4
Equipment Important to Safety for Nuclear Power Plants (AFW
[
q System electrical components, power supplies and instrumentation, and modifications af ter February 22, 1983).
l I
j i
(3) 10 CFR 50 Appendix A.
General Design Criteria for Nuclear i
t i
Power Plants (AFW System).
I J
i (4) 10 CFR 50 Appendix 5.
Quality Assurance Criteria for Nuclear 1
I ii Power Plants and Fuel Reprocessing Plants (AFW System).
i l
(5) 10 CFR 50 Appendix R.
Fire Protection Program for Nuclear Power Facilities operating Prior to January 1, 1979,
(
i Sections III.G. III.J. and III.L (AFW System),
f l
(6) 10 CFR 100.
Reactor Site Criteria (general for Trojan l
Nuclear Plant).
{
{
i I
i 2-2
e e
i i
2.2.2 REGUtATORY OUIDES (1) Regulatory Guide 1.29.
Seismic Design Classification (Rev. 3)
(
september 1978 (AFW System).
i (2)
Quality Assurance Requirements for the i
Installation.. Inspection, and Testing of Instrumentation and i
Riectrical Equipment. August 1972 (APW System instrumentation i
nd electrical equipment).
l t
(3)
Manual Initiation of Protective Actions. October 1973 (APW Syston).
i (4)
Qualification of Class 1E Electrical l
Equipment for Nuclear Power Plants. November 1974 (applicable as described in PCI-1025).
L (5)
Instrumentation for Licht Water Cooled Nuclear Plants to Assess Plant and Environmental Conditions during and following an Accident (Rev. 3). May 1983 (AFV System I
instrumentation).
l l
(6)
Seismic Quslification of Electrical j
Equipment for Nuclear Power Plants (Rev.1). August 1977 (applicable to future modifications per PGE-1028. In-House Position (IMP) No. 1.100-1-2 effective December 31, 1984).
(7)
Regulatory Culde 1.106.
Thermal overload Protection for Electric Motors or Motor Operated Valves (Rev.1) March 1977 (applicable to modifications involving electric motore or motor-operated valves per PCE-1028. IMP Wo.1.106-1-1 effective September 1, 1982).
e 2-3
a e
t (8) Regulatory Guide 1.117.
Tornado Design Classlfication (Rev.1),
April 1978 (applicable as described in PCE-1028, IMP No. 1.117-1-1, effective September 1, 1982),
i h
2.2,3 NURECs i
(1)
NUREG-0578..TMI-2 Lessons Learned Task Force Status Report and l
Short-Tern Recommendations, July 1979.
(2)
NUREG-0588 (category 1).
Interia staff Po** Lion on t
Environmental Qualification of Safety-Related Electrical j
Equipment (AFW System Class 1E equipment purchased before February 22, 1983).
I t
(3)
NUREG-0611. Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidente in Westinghouse-Designed i
Operating Plants.
!j (A) NUREG-0737. Clarifications of TMI Action Plan Requirements,Section II.E.
t (5)
NVREG-0800. Standard Review plan, Section 10.4.9 Auxiliary Feedwater System (PWR), (Rev. 2), July 1981 (used by the NRC for review of the ATW System for SERs).
l
}
2,3 CODES AND STANDARDS
[
1 l
This sec61on lists the codes and standards t.pplicable to the Auxiliary Feedwater (AFW) System.
These sources of codes and standards aret l
(1) American Society of Mechanical Engineers (ASME).
I I
(2)
Institute of Electrical and Electronics Engineers (IEEE).
f l
2-4
s l
l l
l (3) American National Standards Institute (ANSI).
I (4) National Electrical Manufacturers Association (NEMA).
l
- 3.1 CENERAL f
ANSI M45.2.11-1974 Quality Assurance Requirements for the Design of Nuclear Power Plants.
2.3.2 MECHANICAL
[
r (1) ASME Boiler and Pressure Vessel (54PV) Code.Section VIII.
i Division I 1968 and Addenda through 1971 (A W Pumps P-102A and
[
P-1028).
I I
(2)
Draft ASME Code for the Inservice Testing of Pumps in Nuclear Plants. April 1970 (AW Purops P-102A and F-1028).
(3) Draft ASME Code for the Inservice Testing of Valves in Nuclear Power Plants. June 1970 (A W System valves).
(4) ASNI Boiler and Pressure Vessel Code.Section II. 1977 edition.
IWA-7210 (modifications to the AW System).
I (5) ASME Boiler and Pressure Vessel Code.Section XI, 1983 edition and Addenda through the suarner of 1983 (applicable to testing t
of the AW System and components).
(6) ASME Boller and Pressure Vessel Code Section III, Class 2 F
Wuclear Power Plant Components (Contairunent Penetrations 1971
(
and Addenda through Sumer 1971).
t l
(7) ANSI B31.1.0.
Power Piping Code.1973 (Piping and v'alves in the AW System up to the discharge piping, connecting to the Main Feedwater System).
a 2-5
?
6
_m,__.--
e m
f f
I I
2.3.3 ELECTRICAL j
(1)
Criteria for Protection Systems for Nuclear Power Generating stations.ArW System).
(2)
Standard Celteria for Class 1E Power Systems 1
for Nuclear Power Generating Stations (power supplies to ATW System components and instrumentation).
t i
1 (3)
IEE3 Trial Use Standard.
Qualifying Class 1.
4 Electrical Equipment for Wuclear Power Generating Stations (applicable to original ATM Systaa equipment and instrumentation as described in PGE-1025).
E (4)
Standard for Qualifying Class 1E Equiprent for l
Nuclear Power Generating Stations (Applicable to ArW Fystem i
electrical equipment and instrum(ntation as described in i
FCE-1025).
t t
(5)
Standards for the Periodic Testing of Nuc1sar Power Generating Statien Safety Systems (ATW System).
[
1 J
(6)
IEEE Culde for seismic Qualification of Class 1 I
Electric Equipment for Nuclear Power Geneesting Stations (applicable to original ATW System equipment and l
i instrumentation).
l I
t (7)
IEEE Recommended Practices for Seismic Qualifi-
-l cation of Class 1E Eqtipment for Nuclear Power Generating j
j Stations (ATW System equipment ar.d instrumentation modifications af ter 1975).
[
1 (8) NEMA Standard SM-22-1970.
Single-Stage Steam Tur' sines for i
Mechanical Drive Service (K-107A).
i 1
2-6 i
e e
t 3.0 SYSTEM DESIGN SASES
)
This section ytesents desigr. criteria applicable at the systaa level.
Although many criteria are generic in nature, the general and specific requirements applicable to the system are addressed.
3.1 PIPING PRESSURE AND TEMPERATURE CRITIRIA i
1 The piping in the Auxiliary Feedwater (AFW) System is designed according to Power Piping Code ANSI (Amer),can National Standards Institute) 831.1 and the American Society of Mechanical Engineers l
(ASME) Boller and Pressure Vessel Code. All piping within the boundaries of the system, as prescribed in Section 1.2, is constructed t
)
of carbon steel and designed according to the Power Piping Code i
ANSI 131.1, with the exception of the ArW pump discharge piping i
connecting to the Main Feedwater System.
This piping is constructed of l
impact-tested carbon steel and designed according to ASKE Boiler and L
l Pressure Vessel Code Section III Class 2. Nuclear Power Plant Components. Containment Penetrations. The piping classification coding
]
and.,the design, normal and maximum pressure, and temperature ratings
{
l are found in M-301 Specification for Piping Materials and Standard
(
Details, Revision 1.
i
's 3.2 SEISMIC CRITERIA f
CDC 2. "Design Basis for Protection against Natural Phenomena", and Appendix A to 10 CFR 100, "Seismic and Geolot.c Sit.ing Criteria for j
Nuclear Power Plants", require that nuclear power plant structures, components, and systems important to safety be designed to withstand j
the effects of earthquakes without losa of ability to perform their safety functions.
Those structures: components, and systems that have i
3-1 i
i l
l r
i 5
s? < 2 L,'
3 been designed to remain functional in the event of a safe shutdown earthquake (SSR) are designated safety related and as Seismic i
)
Category I (SCI).
The Auxiliary Feedwater (AFW) System design must meet the requirements of CDC 2, and the safety-related portion of the AFW System (the portion necersary for safe shutdown of the plant) must be designed, constructed, and maintained to SCI requirements in j
accordance with Position C.1.s of Regulatory Guide 1.29.
These requirements are included within the scope of IE Bulletins 79-02, f
79-04, 79-07, 79-14, and 80-11, and IE Information Notice 80-21.
The non-safety-related portions are designed in accordance with 3
Position C.2'of Regulatory Guide 1.29.
Seismic response spectra curves for plant areas containing safety-l reltted equipment have been developed for both the operating basis earthquake (08g) and the SSE.
The 08g is assumed to produce a ground acceleration of 0.15 g horizontal and 0.10 g vertical.
The SSE is assumed to produce a ground acceleration of 0.25 g horizontal and 0.17 g vertical.
The entire AFW System is designed to meet SCI requirements with the
'i following exceptionst I
(1) The electric motor-driven AFW pump (P-182).
P (2) The turbine and diese1<. driven pump recirculation lines.
j I
(3) Thw cerdensate storage tank (CST).
I Seismic Category I requirements extend to the first seismic restraint 1
beyor.d the SCI /SCII boundaries, meaning that the first restraint is to l
be a full anchor (restraining six degrees of freedom) where feasible.
j When a full anchor cannot be located within a reasonable distance from i
the SCI /SCII boundary, two 3-way directional restraints are to'oe i
Iccated beyond the boundary.
l I
I r
3-2 I
l f.
L
.n,
.. s:.
e I
l 3.2.1 P-182 l
4
(
j The electtle motor-driven APW pump and associated piping are not
{
safety-related and are not seismically qualified.
This pump is not
(
required for plant safe shutdown because the safety-related portion cf
(
l the A7W System includes two redundant 100 percent capacity pumps.
This classification was faund to be acceptable to the Nuclear Regulatory j
Corsalssion (WRC) in its review for the implementation of WUREC-0611 recossendations for Trojan under TNI Action plan Item II.E.1.1* (see the Safety Evaluation Report (SER) transmitted in NRCato-PCE letter daterd i
Febtvary 18.* 1981).
[
\\
3,2,2 RECIRCULATION LINES i
The recirculation lines from the safety-related pumps to the CST are not seismically qualified downstream of the pressure-reducing orifices 3
\\
u i
and locked-open valves.
A break in the recirculation lines would not l
adversely affect pusip operation.
3.2.3 CST t
The CST (the preferr*d water supply) is not seismically qualified.
The i
backup water supply from the Service Water System to the safety-related AIV pumps is seismically qualified.
Service water is supplied to each i
I pump through a separate SCI piping system.
Each service water line is isolated from the other so that failure of one does not affect the f
other.
Each line is also isolat*d from the SCII CST by nonreturn f
{
- valves, procedures accomplish the switchover from the primary to the t
secondary supply upon loss of the primary supply.
These measures were found acceptable by the NRC in its review for the impleruentation of WUREG-0611 recomendations for Trojan under TMI Action Plan Ites II.E.1.1 (see the SER transmitted in NRC-to-PCE letter dated October 23, 1980).
l t
j 1
J
?
i 3-3 i
k i
p.r i n &
.4'-'Y
I I
I The consson AFV pump suction piping from the CST to the pumps up to the check valves was iniutally classified SCII.
Failure of the suction pipir.3 while the CST water level is above the pump shutdown setpoint could cause the AJV pumps to run dry.
To increase system reliability
(
and prevent this occurrence, this piping was upgraded to SCI downstream I
of the expansion jcint located near the CST nasale as part of RDC 86-002.
RDC 46-002'also added the expansica joint to replace MD-050, the common suction line isolation valve, which was restoved.
The expansiva joint in the coasson suction piping allows telative movement between the CST and piping, and is required for seismic
)
qualification'of the pipe.
[
1 r
The SCI components of the AFV System are identified in the Trojan 1
Q-List. Table VII.
l 3.3 OUAt,ITY ASSURANCE CJt!TERIA CDC 1. "Quality Standards and Records", requires that structures, components, and systems important to safety be designed, fabricated, created, and tested to quality standards consnensurate with the importan** of the safety functions to be performed.
Section 3 of the Updated Final Safety Analysia Report (UFSAR) describes the quality l
l classification system and quality standards that satisfy CDC 1 for i
water-and steam-containing components of the Trojan Nuclear plant.
Group 3 quality standards apply to the Seismic Category I portions of t
l t
the Auxiliary Feedwater (AFV) System. Group 4 quality standards apply to the remaining portions of the AFV system.
Codes and standards used for components and systems specify the design and quality assurance (QA) requirements.
Standards applicable to each quality group are given in Table 3.2-3 of the UFSAR.
For modifications to the design of new systems, the latest effective edition of the l
applicable code, as listed in 10 CFR 50, may be used.
i
~
_ e e>.e,d* _6C)<)
s-... w -
d I M.
e m..
su.
s.
The ATW System at Trojan was designed, constructed. Installed, and tested in accordance with a QA program that meets the requirements of 10 CFR 50 Appendix B.
All modifications are performed in accordance with PCE-8010. Nuclear Quality Assurance Program.
3,4' REDUNDANCY /UTVERSITY CRITERTA CDC 34 "Residual Heat Removal", requires redundancy of Auxiliary Feedwater (AFW) System components so that under accident conditieas, the safety function can be perforned despite a single active compenent failure.
This requirement may be coincident with the loss of off-site power for certain events, in addition. KRC Branch Technical Position ASB 10-1, "Design Culdelines for Auxiliary Teedwater System Pump Drive and Power Supply Diversity for Pressurized Water Reactor Plants", has guidelines that may be used to select the minimum diversity acceptab.'e i
for ATV System pu=p drives and power supplies.
Adequate diversity l
l l
precludes system f ailures due to f ailure of a single type of motive l
I power that can be subject to a failure of the driving component itself, its source of energy, or the associated control system.
Branch Technical Potsition ASB 10-1 specifically recom= ends the following:
(1) The ArW System should consist of at least two full-ca:acity, independent systems that include diverse power sources.
(2) Other powered components of the AFW System should also use the concept of sophrate and multiple sources of motive energy.
The required diversity, for example, might be two separate AFW trains, each capable of removing the r6sidual heat load of the reactor system, with one separate train powered from either of two ac sources and the other train wholly powered by steam and de power.
9 3-5 E * &^ ' ^ ' * -
-r
++
l l
(3) The piping arrangement, both intake and discharge, for each train should be designed to permit the pumps to supply feedwater to any combination of steam generators.
This e
arrangement should take into account pipe failure, active component f ailure, power supply f ailura, or control system failure that could prevent system function. One arrangement that would be, acceptable is crossover piping containing i
i valves that can be operated by remote manual control from the a
control reca, by use of the power diversity principle for the valve operators and actuatie. systems.
i (4)
The AW System should be designed with suitable redundancy to
\\
offset the consequences of any single active component failuret however, each train need not contain redundant i
active components.
f I
(5) With respect to a high-energy line break, the system should
(
l be arranged to assure the ability to supply necessary
[
l j
emergency feedwater to the steam generators, despite the 7
1 postulated rupture of any high-energy section of the system:
a concurrent single active failure is assumed.
I j
The above criteria envelop the generic recomendations identified in h
WREG-0611.
}
1 l
I The AW System at Trojan meets both the redundancy requirements of i
CDC 34 and the guidelines of granch Technical position Ass 10-1.
The system is designed with adequate redundancy to accomodate a single
{
active component failure without loss of function.
Diversity in pump motive power sources and essential instru..entation and control power j
j sources is provided.
i t
f i
a 3-6 f
I
.~
- m. A m.....a
-:..A
I The A N System provides two redundant and independent means of supplying feedwater to the steam generators for cooling the RCS under emergency conditions.
Two full-capacity A W pumps are provided with diversity in drivers.
pump p-102A, in AW Train A, is driven by a steam turbine.
Pump P-1028, in AW Train B is driven by a diesel engine.
There are no valves in the cornon piping to the condensate storage tank (CST) to ' prevent AW pump damage due to operation without water.
The discharge piping of each A N pump is separated from the other by motor-operated isolation valves and check valves until they
, join into a single line before connecting with the main feedwater line of each steam generator.
Each steam generator is separately' supplied with AW through its main feedwater line.
The pumps take buction from two sources.
The normal source is the SCII CST.
If tMs source f ails, water can be supplied to the punps from the SCI Service Water System (SVS).
Service water is supplied to each pump through a separate SCI piping system.
Each 3*.'s supply line is isolated from the other so t. hat failure of one does nut affect the other.
Each service water line is also isolated from the CST by nonreturn valves, l
An SCII, full-capacity AW pu=p (p-182) is used during normal startup and shutdown to reduce wear on the SCI ESF AW pumps.
In an emergency, this pump may be aligned to AW System Train A or B should both SCI pu.ps fail.
This capability to provide inc. eased AW System availability extend:: beyond normal single failure criteria and thus provides increased reliability. However, operation of AW Puep p-182 is not assumed for any accident analysis.
Complete physical and electrical separation is maintained throughout the pump control, control signals. electrical power supplies, and instrumentation for each safety-related AW pump.
Class IE electrical components in the AW System are powered by the Engineered Safety Features (LSF) Electrical Distribution System.
A W System Trains ' an'.
B are powered by ESF Instrumentation Channels A and B.
SCII AW Purp p-182 may be manually loaded on either 4.16-kV ESF bus if needed.
3-7
,,, j. 2 m o oce n
~.
l i
t i
i i
1 b5 ENVIRONMENTAL QUALIFICATI0W CRITERIA The anvironmental qualification (EQ) review of electric equipment important to safety is perfomed in accordance with 10 CFA 50.49.
[
I CDC 4. "Environmental and Missile Design Bases", establishes the l
J general requirement for environmental qualification of safety-related equipment.
It states in part that "structures, systems, and components important to safety shall be designed to accoeunodate the effects of and to be compatible with the environmental conditions associated with j
normal operation, maintenance, testing, and postulated accidents, including loss of coolant accidents".
i l
IEEE 323-197a. "!EEE Standards for Qualifying Class 1E Equipment for Nuclear power Generating Stations", is the current industry standard I
for environmental qualification of safety-related electrical I
l equipment.
This standard was first issued as a trial-use standard.
IEEE 323-1971. The 1974 standard includes specific requirements for j
aging, margins, and document maintenance that were not included in the l
1971 trial-use standard.
The 1974 standard was endorsed by the Nuclear
{
}
Regulatory Comission (FRC) in Regulatory Guide 1.F9 for plants with i
existing construction pomit applications.
t j
Design of the Trojan Nuclear plant was initiated before the issuance of I
f 4
IEEE 323-1971. PCE comitted to the Westinghouse Qualification
(
i i
program, which used IEEE 323-1971 as a design standard for many Nuclear
{
l Steam Supply System (NSSS) corponents, especially those inside t
4 Containment.
l In 1979, the Division of Operating Reactors (DOR) published "Culdelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors", containing definitive criteria for reviewing the EQ of safety-related electrical equipment.
The' intent of the D0R guidelines is to prcvide a basis for judgments required to I
a 4
3-8 i
i a
~.sd - aOS
-c
-s
~ *
- t------
I i
l l
confirm that operating reactors are in compliance with CDC 4.. It i
f Provides a number of VRC staff positions on selected troas of the i
qualification issue. Category 1 positions apply to equipment qualified l
in accordance with IEEE 323-1974. Category 2 positions apply to i
a equipment qualified in accordance with IEEE 323-1971.
In June 1984 j
i the NRC issued Revision 1 to Regulatory Culde 1.89. which describes a t
method acceptable to the NRC staff for complying with 10 CT.T 50.49.
[
The guide endorses IEEE 323-1974, subject to certain :larifications and conditions.
The D0R widelines. WREC-0588, and to CTR 50.49 are implemented at Trojan in accordance with the followingt Equipment ordered prior. to February 22, 1983 is qualified in accordance with D0R
'l l
guidelines or WREC-0588 (Category 1).
Equipment ordered af ter February 22, 1983 but before June 1. 1984 is qualified in accordance j
with WREC-0588 (Category 1), unless special provisions are applicable as delineated in pCE-1025.
Qualification in accordance with WREC-0588 is construed to be equivalent to meeting the provisions of l
j 10 CFR 50.49.
Equipment ordered after June 1. 1984 should be qualified in accordance with 10 CTR 50.49. unless special provisions are applicable.
The provisions of Revision 1 to Regulatory Guide 1.89 are complied with in the qualification review of this equipent, except as i
noted in pCE-1028. Regulatory Guido policy Manual.
[
I j
The EQ program also includes electric equipent not currently in the j
scope of 10 CFR 50.49, including electric equipent important to safety
{
i I
in mild environments and certain postaccident eenitoring equipment in
[
l harsh environments specified in category 3. Regulatory Guide 1.97.
These equipent categories are included in the EQ Program to ensure
(
l that all equipment qualification efforts are effectively integrated I
1 throughout plant design.
l r
)
P L
The AW e,o:eponents important to safety in harsh environments aret J
j j
(1) Various terwinal boards in the electrical penetration and
[
main steam support structure (MSS).
i t
i i
3-9 i
i i n,..
.w e ?
l P
i l
l (2)
The four turbine steam supply solenoid valves and associated 11 alt switches.
L (3) The eight CV-3004 series flow control valves and associated j
flow indication switchen.
1 j
(4) Yne eight 3043, series flow transmitters.
)
(5) Power, instrumentation, and control cabling.
The Engineered Safety Features (ESF) AN pumps are located in a alld j
environment in the Turbine Buildir.g. Coolers have been provided on 5
K-107A drain lines to reduce humidity, and enclosures for three SCI J
components have been provided with additional sealing to increase reliability because of high humidity in the F-102A pump room.
Supply i
and exhaust fans for each EST AW pump room automatically start upon a j
pump start.
r 1
r 1
2
]
High energy line breaks in the AW pump rooms were excluded from
]
consideration as described in PCE-1004, because the redundant trains
)
are separated frca each other, as well as from the effects of a main steam er feedwater line break, i
l Table 3-2 of PGE-1025 is the master list of electrical equissent l
important to safety at Trojan within the scope of 10 CFR 50.49(b) and i
is generated from a sort of the Component suasaary Sheets (CBS) for 1
equipment in harsh environments. The identification of equipment important to safety also documents electrical equipment important tc safety in al14 environttents.
i Table 4-1 of PCE-1025 lists the normal and accident environment i
conditions for each plant location.
i 1
l I
i i
3-10 4
e
+ ' * '
.-OP'
3.6 FIRE PROTECTION CRITERIA CDC 3. "Fire Protection", requires that structures, systems, and companents isrportant to safety be designed and located to minimize the probability and effect of fires and explosions.
10 CFR 50.48(a) furthee requires that the Trojan Nuclear Plant implement a fire protection program.
The Auxiliary Feedwater 3/ stem is a safe shutdown system per 10 CFR 50 Appendix R.
PCE-1012. "Trojan Nuclear Plant Fire Protection Plan", describes the fire protection features at the Trojan Nuclear Plant that implea.ent compliance with the above requirements.
The NRC SERs dated March 9, 1978. March 25. 1980 and October 6.1980 document NRC acceptance of features meeting requirements of Branch Technical Position (BTP)
APC359.5-1 Appendix A. Guidelines for Fire Protection for Nuclear 3
Power Plants Docketed Prior to July 1. 1976. This has since 5een incorporated in BTP CMEB 9.5-1. Guidelines for Fire Protection for Nuclear Power Plants and CDC 3.
Also, an NRC SER dated October 15 1985 documents NRC acceptance of features meeting the requirements of in CTR 50. Appendix R. Sections III.C.3 and III.L.
These requirements, and NRC-granted exceptions, fom the basis for the Trojan Nuclear Plant fire protection desigr. and are described in detail in PCE-1012.
3,7 E?NTR0!!MNTAL PROTFC"?f 0?l CRITFRI A CDC 2. "Design Bases for Protection Against Natural Phenottena",
requires that stmetures, systems, and components teportant to safety be designed to withstand the effects of natural phenomena such as l
earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of ability to perfom their safety functions.
CCC 4 l
"Environmental an4 Missile Design Bases". requires that st mctures.
systems, and components ieportant to safety be designed to c:comodate the effects of and be corpatible with the envirom. ental conditions associated with nomal operation, maintenance, and testing and with 3-11
(
L w..
.m.
postulated accidents.
It also requires that they be appropriately protected against dynamic effsets. including missiles, pipe whips, and discharging fluids.
This section describes those criteria addressing i
flooding protection, missile protection, tornado protection, pipe whip.
and jet impingement.
3.7.1 FLOODINC PROTECTf0Pi The design basis for flooding protection of the ADI considers two types of floodtog events:
external flooding, and internal flooding due to tvpture of fluid system piping.
3.7.1.1 External Plooding The design bati: natural flooding event for the Trojan site is the i
probable maxteum flood (pMT) for the Columbia River, established by the U.S. Army Corps of Engineers.
The maximum water levet at the Plant site for the pMr is 39.2 f t mean sea level (Mst), based en a flood j
elevation of 36 it above MSL corresponding ~to a river dischatgo of 2.2 million cubic feet per second (cts), plus a wave run-up of 3.2 ft.
t The damign basis artificial flood for the Trojan site is a flood i
resulting from a catastrophic failure of the Crand Coulee Dam due to the massive effects of a nuclear weapon.
This postulated event would
[
result in a short-duration river discharge of 3.6 million to l
4.4 million cfs at the Trojan site approx!r4tely 2 days af ter dam failure.
The maximum flood elevation for 4.4 million efs. on the basis l
of artificial flood hydrographs calculated energy slopes, and 1
Manning's foreula, is 41 f t MSL. plus a wind wave run-up of 1.75 f t.
i for a water level at Plant site of 42.75 ft.
l The floodir.g protection criterion is leplerented at Trojan by, virtue of l
the f act that all corTonents of the AIV System are at least 45 f t above MSL. and therefore above any credible flood level.
I l
l 3-12 4.e.
- .1
o L 7.1.2_ _ InterMLFliodins The design basis internal flooding event for the AW System is a massive circulating water pipo rupture in the Turbine Building.
The assumed maximum flood rate is based on a design flow rate of 210.000 spm for each of the two Circulating Water system pumps.
1 Protection for the design basis event is laplemented by the flood relief dampers on the west well of the Turbine tuilding and a concrete and steel dike that surrounds the SCI AW pumps and remote shutdown station rooms.
The 2-ft-high flood relia'f dampers are assumed to accospodate a discharge rate of 500.000 spa to the plant yard.
Therefore, a 2-ft-high dike is sufficient to prevent flooding of the SCI AW pumps and remote shutdown station rooms.
Orain line protectica is provided in the rooms to prevent backflooding of circulating water into the rooms.
3,7,2 MisstLE PROTECTICW As required by CDC 4 the AW System must be appropriately protected against the dynamic effects of missiles. The design basis turbine missiles are those from the last-stage wheel of the low-pressure turk.tne of the turbine generator, and a bucket vane from the low-pressure turbine of the turbine generator. Maximum assumed shield penetration would result from a low trajectory (predominantly horizontal) missile from the 120-degrt'e segment of the last-stage wheel of the low-pressure turbine, which would penetrate 3 ft 0 1. into 3.000 psi reinforced concrete, or 1 ft 9 in into 6.000 psi reinforced uncrete.
Protection from turbine missiles is implemented through the combined thickness of slabs above the AFW pumps.
No other internally generated missiles are assumed to present a significant hazard to the AW System.
The SCI portion of the A N pump suction piping located outside the Turbine Building is protected by a concrete slab.
3-13
- d _, e> 1_ _
"s*-
4*
y...
.4
o I
t 3.7.3 TORFADO PROTECTION l
As required by GDC 2, the AFW Systen must be designed to withstand the effects of tornadoes without a loss of ability to perform its design I
safety function. The design bases reflect appropriate consideration of the.sont sevore of the natural phenomena reported for the site, i
appropriate combination of the effects of normal and accident conditions with the effects of the natural phenomena, and the importance of the safety functions to be performed.
The design sind velocity is 105 mph 30 ft above ground elevation of 45 ft above M31..
The design basis tornado load for the ADI pump rooms is a 200-eph tornato.
This load is basei on analysis per American l
Society of Civil Engineers (ASCL) paper 3269. Wind Forces on Sttvetures, and reflected in the Trojan Nuclear plant SgR dated October
]
1974 The design basis missiles for the 200-mph tornado are equivalent I
to
{
l
(
(1) A 4 iti. by 12 in. by 12-fL long wood plank traveling and on
(
)
at a velocity of 200 mph at any elevation of the strveture, i
1 5
l (2) A 3 in. diameter by 10-ft long ASA Schedule 40 steel pipe l
traveling end on at a telocity of 75 mph at any elevation of
{
the sttveture.
3 I
(3) A 4000-1b passenger car striking with a ve19etty of 40 rph f
I not more than 25 ft above grade.
I
(
i i
Combined tornade loading takes into account tha wind loading. the j
tornado-induced pressure differential, and tornado-generated missiles to produce the most critical loading condition.
I t
I j
The SCI portions of the AIV System are protected against tohadoes and tornado missiles.
Exposed sections of AIV pump suction piping and 1
i i
}
a t
i f
3-14 i
i t
3.. 3,
.e.
.w As i
L t
condensate storage tank (C3T) level instrumentation located outdoors are proteci.ed by enclosures.
The CST level instrumentation is required l
to provide level indication in the control room and the remote shutdown _
station, and to stop the SCI A W pumps upon a low CST level.
4 The redundar.t level transmitters are on opposite sides of the CST and are l
provided with missile-proof enclosures. The output of the transmitters "f ails low" to trip the. AW pumps if the transmitters break aw.ny frora i
the CST.
'T' 3,7.a PIPS WHIP AND Jg7 IMPINCDfgNT As required by CDC 4. the AW system design must include protection I
i t
against dynamic effects, including the effects of pipe whipping and l
discharging fluids that may result from equipment failures and events i
and conditions outside the nuclear power unit.
(
Bechtel Power Corporation Topical Report BN-70P-2. "Design for Pipe j
Break gffects" 4escribes analytical techniques to analyze fluid jet i
impinsement loads and to establish design criteria for pipe restraints provided to prevent damage from pipe whip effects resulting from postulated pipe breaks in nuclear power plants.
Pog-1004. "Trojan i
Nuclear Plant Analyses of pipe Systen Breaks. Outside Containment".
]
describes the analysis perfnrmed to verify the acceptability of plant design.
High-energy lines are defined as piping systems whose i
operating temperatures exceed 200'P and/or whose operating pressure a
j i
exceeds 275 psig.
I l
The AW System lines analysed in PCE-1004 are the steam supply lines to
}
}
the ATW turbine driver and the A N discharge lines to the steam j
generators.
1 l
Request for Design Chang,e (RDC)80-054 added a guard pipe around the
]
turbine-driven AW punp discharge line in the diesel-driven pump room
{
}
so that a tvpture in a turbine pump discharge line would not damage the f
i 4
i j
3-15 I
t
-D5
- M'
O diesel-driven AFW pump.
This RDC was based on recommendations from KRC-to-PCE letters, dated October 3. 1979 and May 14, 1980.
RDC 80-061 connected the discharge of the motoradriven AFW pump to the discharge of the turbine-driven ATW pump to improve system reliabl%ity in accordance with NUREC-0737 and recommendations from NRC-to-PCE letters dated Octocer 3.1979 and May 5.1980.
3,8 AVT0MATIC IWITIATION GDC 20. "Protection Syrtem runctions" requires that a protection l
system be designed to initiate protective action automatically to assure that acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences.
CDC 34 requires that the safety function of the designed system, that is, residual heat removal i
by the Auxiliary Teodwater (AFW) System, can be accomplished even it, the case of a single failure.
In addition, to improve the reliability of the ATW System. lic6nsees are required to upgrade the system to ensure timely automatic initiation when necessary.
The recommendations
(
of NUREC-0578. Section 2.1.7.a. ares l
(1) The design shall provide for the automatic initiation of the ATW System.
(2)
The automatic initiation signals and circuits shall be designed so that a single failure does not result in the loss of ArW System function.
(3)
Testability of the initiating signals and circuits shall be a
{
feature of the design.
(4)
The initiating signals and circuits shall be powered from the emerger.cy buses.
3-16 e i p c..
.O i
(5) Manual capability to initiate the AfV system from the control room shall be retained and shall be implemented so that a single failure in the manual circuits does not result in the loss of system function.
(6) The AC notor-driven pumps and valves in the AIV System shall not be included in the automatic actuation (simultaneous and/or sequential) of the loads to the eftergency buses.
(7) The automatic initiating signals and circuits shall be designed so that their failure does not result in the loss of manual capability to initiate the AJV System frcm the control room.
In the long term, these signals and..ircuits are required to be upgraded in accordance with safety-grade requirements.
Specifically, in addition to the items itentioned above, the following is required autematte initiation signals and circuits must have independent channels, use environ:centa11y qualified components have systes
[
bypassed / inoperable status features, and conform to control system interaction criteria. as stipulated in IEEE 279-1971.
The initiation signals logic, and associated circuitry of the Trojan Nuclear Power plant AFV autcmatic initiation system comply with the long-tem safety-grade requirements of WUREG-0578. Section 2.1.7.a. and the subsequent clarification issued by the Nuclear Regulatory l
Comission (KRC), with the exception that the initiation circuits for loss of both main feedwater pumps are not cafety grade.
The NRC has detetulned that because the trip of both main feedwater pmeps is a i
secondary AIV system automatic initiation signal for Westinghouse f
plants and because there are other diverse safety-grade AJV System l
automatic initiation signals provided in the Trojan design, th.e I
circuitry providing initiation on the tripping of both main feedwater 3-17 l
(
a e4 !..a.es e + + -
o l
l r
pumps does not have to comply with safety-grade requirements (see 1
FRC-to-p0E letter dated April 8, 1982).
No credit is taken for this initiation signal in the accident analyses.
i At Trojan, logic channels supply initiation signals to the control circuitry and automatically initiate /A flow under any of the following conditions:
l t
(1) Turbine-driven MW pump.
l l
(a)
Safety injection (Train A).
(b) Steam generator (30) low-low level (two of three level transmitters in any one SG).
(c) Undervoltage on 4.16-kV Sus A1.
(d) Trip of both rain feed pumps.
I (2)
Diesel-driven MV pump.
f (a) safety injection (Train 5).
1 l
(b) 30 low-low level (two of three level transali,ters in any
[
one SG).
i (c) Undervoltage on 4.16-kV Dus A2.
(d) Trip of both main feed pumps.
(
t Section 4.17 of IEEE 279-1971 requires that protection systems include f
means for manual initiation of each protective action at the systen level and that the single-failure criterion as set forth in Section 4.2
(
of IEEE 279-1971 be set.
I l
Manual operation of the MV System is provided in the Ccntrol Ronm and f
at Panel C-160.
Each control circuit is independent so that a single failure in one train does not uffect the redundant train.
In, addition, the automatic initiating circuits are designed to be electrically
[
independent from the Control Roon manual start circuit so that failure t
3-1:
... < e> u.f ; O ___._. _ _.. a.__
c-y,:
a 6
of the automatic initiating signals does not affect the control room manual capability of AFW pu=ps.
Manual initiation may be accomplished at the following locations:
(1) Turbine-driven AFV *, an.
(a) Local (C-160).
(b) Control room.
(2)
DieFel-driven AFV pump (a)
Local (C-160).
(b) Control room.
(3)
Motor-driven AFV pump.
(a)
Local (at the 4.16-kV A5 switchgear).
(b)
Control room.
System valves are aligned for normal operation (normally open);
therefore, flow is initiated upon the startup of either pump.
The operation of either automatically initiated pump provides the capacity to remove decay heat from the steam generators at a rate sufficient to prevent overpressurization of t3e heactor Coolant Syrtom (RCS) and to maintain steam generator levels.
Initiation of AFW flow within 1 min.
of receipt of an automatic start signal provides sufficient capacity to remove decay heat, prevent overpressurizing the RCS. and provent uncovering the reactor core under the postulated accident conditions.
This condition was an assumption of accident analysis (see Section 1.3) and thus no margin may be assumed.
Consequently, the AFW System is capable of automat ally initiating appropriate protective action, with precision and re116;ility, whenever a condition monitored by the system
~
reaches a preset level.
3-19 a : 4 A.'I. Vk&2 : w. o eh r~as. =
t s i s.' A
>a;
.< s
The protection system is designed to be independent of the control system.
In certain appilcations, the control signals and other nonprotective functions are derived from individual protective channels through isolation amplifiers.
The design meets the requirements of GDC 34.
The automatic initiation signals and circuitry for the AFW System at Trojan comply with the single-failure criteria of IEEE 279-1971.
The Trojan Plant has the ability to test the AFW initiation system while at power.
The automatic start circuits are designed with provisiona for both periodic testing and calibration during normal plant operations (see PGE-to-NRC letter dated October 17, 1979).
Trojan's 31-day and 92-day interval Periodic operating Tests (POTS) 5-1 and 5-2 and 18-month interval POTS 5-3 and 25-2e outline the methods and reporting procedures to be used in testing the pumps, valves, and their initiating circuits required by the Technical Specifications (Paragraphs 4.7.1.2.1 and 4.7.1.2.2).
All locked valves (FW 081-094 FW 103-114 and FW 119-120) in the auxiliary feed flow path are checked during the monthly POT.
The auto-start signals for the steam turbine-driven and diesel-driven AFW pumps are powered from 120-V preferred instrument"ac Buses Y11 and Y22. respectively.
e 3-20
- 4..
a w 2: 5: a % : L'"'
'M S -
4.0 COMPONENT DESIGN BASES This section consolidates and addresses information regarding design basis requirements for individual (or groups of) components within the system.
Discussion concentrates on component-level design criteria, which may or may not relate directly to the system performance requirements discussed in Section 3.0.
Where appropriate, earlier discussions are cited to avoid repetition.
4.1 AFW PUMPS P-102A and P-102B The two engineered safety features (ESP) auxiliary feedwater (AFW) pumps in the AFW System automatically start to supply feeu.ater into the steam generators as required in Section 1.3.4 The design requirements concerning AFW Pumps P-102A and P-102B are described below.
4.1.1 DESIGN REQUIREMENTS (1) AFW Pumps P-102A and P-102B must be quallfled as SCI components per Section 3.2 as required by Position C.1.s of Regulatory Guide 1.29.
(2) The two AFW pumps. P-102A in Train A and P-102B in Train B, must meet redundancy and single-failure criteria as described in Section 3.4.
(3) The AFW pumps must have diverse drivers to meet the criteria stated in BTP ASB 10-1 relating to CDC 34.
AFW Pump P-102A is driven by Turbine Driver X-107A. and AFW Pump P-102B is driven by Diesel Driver K-1078.
(4) The net positive suction head (NPSH) required at the suction of the AFW pumps at the design flow rate of 960 Spm is 4-1
- x k ihi+9 W~~.i**s~ \\&%A
~.
25 ft.
The required FPSH is established by the pump supplier and verified in Bechtel Calculation 16-42.
Bechtel Calculations 16-44 and 16-47, and Nuclear Plant Engineering (NPE) Calculation TE-115 also address NPSH cn the basis of the condensate storage tank (CST) level during both one-and two-pump operation.
(5)
AFW Pumps P-102A and P-102B must meet design pressure and temperature ratings specified in Section 3.1.
(e)
AFW Pump P-10?A must operate independently of ac power as required by Recommendation CL-3 of the NRC-to-PCE letter dated October 23, 1980.
4.1.2 COMFIGURATION The two ESF AFW pumps are identical six-stage, horizontal, centrifugal pumps.
Both are equipped with radial, pressure-lubricated bearings, which are journal bearings with a Kingsbury thrust bearing on the outer end.
Bearing lubrication is supplied by a shaft-mounted DeLaval oil pump.
The lubricating oil is cooled by an internal lube oil cooler.
Cooling water for the P-102A lube oil cooler is provided by the pump itself to allow operation independent of ac power.
Service water was ured to' cool the lube oil cooler of P-102A until RDC 78-020 modified the cooling water flow path to allow pump discharge water to cool the lube oil.
Service water can be used as a backup source of lube oil cooler cooling water flow.
For P-1028, cooling water flow to the lube oil cooler is supplied from the Service Water System.
Mechanical seals are located on each end of the pump shaft.
The seals are cooled by a portion of the pump discharge flow, which is then returned to the CST.
A minimum flow orifice is located in each of the AFW Pumps P-102A and P-102B recirculation lines to prevent overheating 4-2 w,,..
, u i..i. &,.
4 6 'd e
of the pump during minimum flow operation.
The orifice is rated for a flow of 80 spa (minimum recirculation flow rate) with a pressure drop of 1,470 psi.
Applicable AFW Pumps P-102A and P-102B design and operating data provided by the pump supplier, Bingham Willamette Company, are listed in Table 9-1.
4.1.3 MARGIN EVALUATION Design m&rgin for AFW flow rate for accident analyses is held by Westinghouse and is not available to PCE. However, the specified pump ratings of 880 spm exceed the minimum flow rate of 426 gpm assumed for the worst-case accident analysis.
The pump supplier minimum recirculation flow is 80 spa, and no margin is provided.
The margin between available MPSH (38.3 ft) and required MPSH (25 ft) is 13.3 ft as verified by Bechtel calculation 16-42.
4.2 TURBINE DRIVER K-107 A I
Turbine Driver X-107A receives steam from each main steam header to drive Auxiliary Feedwater (AFW) Pump P-102A.
It automatically starts to mechanically power P-102A to send feedwater into the steam generators as required in Section 1.3.4 4.2.1 DESICM REQUIREMENTO (1) Turbine Driver X-107A must be qualified as a Seismic Category I component per Section 3.2, as required by Position C.1.s of Regulatory Guide 1.29.
(2) The turbine driver must be capable of automatic startup as regulred by CDC-20 and acceleration to deoign speed 'within I
the required time (60 seconds) to meet the required flow conditions per Section 1.3.4 4-3 r3n.
- u. Js.4 h.. #;.a:.~. u a,..-d
./
~
(3) The control power for Turbine Driver K-107A must be from a class.1E' 125-V de power source via inverters to ensure that P-102A operates independently of ac power for at least 2 he as required by Recommendation CL-3 of NUREC-0611 (see the NRC-to-PCE letter dated October 23, 1980).
(4) Turbine Driver X-107A operates on steam pressure supplied between 125 psia and 1,220 psia (lowest setpoint of main steam safety valves with full accumulation).
(5)
Sufficient and redundant instrumentation and controls must be located in the control room and at the Remote Shutdown Panel (C-160) to allow for proper operation and monitoring of turbine Driver X-107A as required by CDC 13 and CDC 19.
4.2.2 COMFIGURATION Turbine Driver K-107A is a single-stage, noncondensing turbine mechanically coupled to AFW Pump P-102A.
Turbine Driver K-107A is located within a reinforced concrete enclosure at the east end of the 45-ft level of the Turbine Building.
X-107A automatically starts and accelerates to 4,560 rpm within 60 seconds if any of the following start signals 19 present:
(1)
Safety injection signal Channel A.
(2)
Undervoltage on 4.16-kV ESF Bus A1.
(3)
Low-low water level (11 percent) from two of three level transmitters in any steam generator.
(4)
Loss of both main feedwater pumps (2/2 coincidence).*
i l
4-4
+..
- ~ ~ +
~
l Steam is supplied to K-107A from each of the four main steam headers.
The suppih lines tap off each header upstream of the main steam isolation valves.
Steam is supplied to K-107A from each main steam header through an air-operated control valve (CV-1451 through CV-1454). The four supply lines converse to a common header containing MO-3170 (steam stop valve) and HO-3071 (turbine trip and throttle valve).
The air-operated control valves are normally closed, but the steam lines downstream of them are kept warm by a small amount of steam flowing through bypass orifices.
These orifices allow steam flow to K-107A when it is idle to ensure a quick starting capability.
Drain coolers, cooled by service water, are installed on lines draining condensed steam to reduce the humidity in the p-102A pump The steam pressure required to operate X-107A was verified by room.
PCE Calculation TM-006 in response to License Change Request LCR-76-16.
Controls for operation of X-107A are located in the control room and locally at panel C-160.
Normally, speed is automatically controlled by a governor that compares a required differential pressure signal between the p-102A discharge pressure and the main steam pressure to an actual differential pressure signal.
This feature is described in Section 4.11.
The Woodward Electric Governor System provides reliability to prevent overspeed upon startup.
Overspeed trips protect both the pump and the driver, and limit the pump discharge pressure to the maximum allowable by the piping design (Section 3.1).
The electrical trip is sot at 5,100 rpm to protect the machine fram damage and limit the discharge pressure of the pump.
PGE Calculation TM-002 verified that no damage would occur'with p-102A j
operating at just under the original electrical trip setpoint of l
5,480 rpm.
The mechanical overspeed trip, set at 5,500 rpm, a backup for the electrical overspeed trip, also prevents equipment damage.
When testing the mechanical overspeed trip. Turbine Driver X-107A is uncoupled from AFW Pump p-102A to prevent overpressurizing the pump I
discharge piping.
i i
I I
I l
4-5
. ce,,c j W Ae. <. _. < m__'.a.
n
l A level switch trips the turbine driver upon a 30 percent water level in the condensate storage tank (CST).
The trip setpoint was determined as part of PGE Calculation TE-115. AFW Pump / CST Level Control Modification for Request for Design Change (RDC)84-096.
An override allows operators to manually restart X-107A at a lower AFW System flow rate if it was shut down on t'.ie CST low-level trip.
Control power for the K-107A Woodward electric governor is supplied from 24-V de from JQ-2079A in C-160 which is supplied from a Class 1E 120-V ac power source panel Y22.
X-107A is designed to operate for at least 2 he independently of ac power.
This compliance to NUREG-0737 Item II.E.1.1. is verified by M6morandum ANR-172-84. which cites the following:
(1)
PGE Calculation TE-028:
Evaluated the station battery capacity under various temperature and battery age conditions.
(2) PGE Calculation TE-029:
Evaluated the load shed necessary to maintain an operable station battery for a minimum of 2 he after a loss of ac event.
(3)
E. L. Davis to H. E. Williams Memorandum ELD-03-82M. dated December 21. 1982:
Summarized the station battery capabilities and recommended load shedding from the A train in order to meet the 2-hr requirement.
(4)
R. L. Steele to C. P. Yundt Memorandum RLS-03-83M. dated l
January 4 1983 Recommended a procedure change to maintain an operable A train battery and AFW pump for 2 he after a station blackout.
(5)
C. P. Yundt to R. L. Steele Memorandum CPY-443-83 dated June 15. 1983:
Documented completion of the Nuclear' Plant Engineering (WPE) recommended procedure change.
4-6 i
.?.
n e..h O u-6.b
..u
Cooling water to the turbine lube oil cooler is supplied from P-102A.
This flow keeps the K-107A lube oil temperature between a minimum of 110*F (cooler outlet) and a maximum of 185'F (cooler inlet).
Applicable Turbine Driver K-107A design and operating data provided by the turbine supplier Terry Steam Turbine, are listed in Table 9-2.
4.2.3 MARCIN EVALUATION The design steam inlet pressure range of 125 psia to 1,320 psia provides a margin that envelops the steam pressure range of 125 psia to 1,220 psia, at which the AFW System is designed to operate.
The maximum K-107A inlet steam moisture condition of 0.5 percent by weight is twice the design moisture content of steam exiting the steam generators.
Tha internal steam generator. moisture separation equipment is designed to ensure that moisture carryover doe s not exceed 0.25 percent by weight.
4.3 DIESEL DRIVER X-107B Diesel Driver K-107B is designed to automatically start and mechanically power P-102B to supply feedwater to the steam generators as required in Secticn 1.3.4.
4.3.1 DESIGN REQUIREMErlTS j
(1)
Diesel Driver K-107B must be qualified SCI per Section 3.2 as required by Position C.1.g of Regulatory Guide 1.29.
(2)
Diesel Driver K-107D must be capable of automatic startup as required by CDC 20 and acceleration to design speed within the required time (60 seconds) to meet the required, flow conditions per Section 1.3.4 i
4-7 i
0
- n..%
cs.
y th*.m W.ee 4.* GQL
\\
i i
(3)
Control power for the diesel driver is from the Diesel Driver K-107B nickel-cadalum 28-V battery.
y (4)
Sufficient and redundant controls and instrumentation must be located in the control room and at the Remote Shutdown panel (C-160) to allow for proper operation as required by CDC 13 and CDC 19.
t (5) The auxiliary feedwater (AFW) pump diesel fuel oil day tank l
must provide sufficient fuel to operate K-1078 for 10 he t
under design flow conditions (960 spa at 3,400-ft head).
L 3. 2 CONTICURATION i
l Diesel Driver K-1078 is a 12-cylinder Waukesha diesel engine, supplied with No. 2 diesel fuel and mechanically connocted to p-1028 through a
[
speed increaser (X-173).
The speed increaser ratio of 1:3.8 allows I
K-1078 optimum speed (1,200 rpm) to match p-102B speed (4,560 rpm). A I
12-cy11nder diesel engine is used to develop a maximum-rated horse-i power of 1,579 bhp at 1,200 rpa to ensure proper operation of p-1028.
K-1078 automatically starts and accelerates to 1,200 rpm within 60 seconds if any of the following start signals is presentt i
(1)
Safety injection signal Channel B.
[
f (2)
Undervoltage on 4.16-kV engineered safety features (ESP) i i
Bus A2.
i I
(3)
Low-low water level (11 percent) from two of three level J
transmitters in any steam generator.
l 1
)
l (4)
Loss of both main feedwater pumps (2/2 coincidence)/
j i
l 4-8 i
- "'5" O
"'"*W*
s,-
- 4.
t Q
Diesel fuel is gravity fed from T-152 the AFW pump diesel fuel oil day tank.
The tank has a capacity of 500 gallons.
The Technient Specifications require at least 450 gallons of fuel in T-152 for P-102B to be operable.
This level has been verified sufficient by Bechtel calculation 12-22.
Controls for the operation of K-107B are located in the control room and at Panel C-160.
Normally, speed is automatically controlled by a governor that compares a required differential pressure signal between P-102B discharge pressure and main steam pressure to an actual differential pressure signal.
This feature is described in Section 4.11.
A Woodward electric governor system prevents diesel engine overspeed during strartup.
An overspeed trip device trips K 'A07B if engine speed ey.ceeds 1,350 rpm to protect the diesel engine.
The setpoint of the trip ensures that pump discharge pressure does not exceed allowable limits as specified in Section 3.1.
This protection has been verified by Bechtel Calculation 16-52.
The quick starting capability of X-107B is also enhanced by jacket water system heaters to keep jacket water warm and a soak back pump to circulate lube oil when K-107B is idle.
A level switch trips K-107B at 35 percent water level in the condensate storage tank (CST).
The trip setpoint was determined as part of pCE Calculation TE-115 AFW Pump / CST Level Control Modifica-tion for ROC 84'-096.
An override allows operators to manually restart X-107B at a lower ArW System flow rate if the pump shuts down on the CST low-level trip.
Both control and starting power for X-107D are supplied from a Class 1E dedicated 24-V battery.
The Service Water System supplies cooling water to the diesel. lube oil cooler, intercooler, jacket water cooler, and speed increaser' gear lube oil cooler.
Service water flow to X-107B and X-173 is normally 4-9
.xJ ' r iwaM -
' att W '3'
isolated, but when any of the four Y-10?B automatic start signals is Present, HO-3060B opens to initiate cooling water flow.
Applicable X-107B design and operating data provided by the diesel supplier, Waukesha, are listed in Table 9-3.
4.3.3 MARCIN EVALUATION The margin between actual T-152 capacity of 500 gallons and the volume required to operate P-1028 for 10 he under design flow conditions (960 spm at 3,400-ft head) has not been verified.
4.4 AW PUMP P-182 Auxiliary feedwater (AEW) Pump P-182 is a non-safety-related pump.
It is used to supply feedwater to the steam generators during normal plant shutdown and startup.
It can also be used as a backup to the two engineered safety features (ESF) AFW pumps. P-102A and P-1028.
4.4.1 DESIGN REQUIREMENTS (1) Pump P-182 must meet design pressure and temperature ratings as specified in Section 3.1.
(2)
Pump P-182 must be capable of meeting system design flow requirements if both safety-related pumps are inoperable.
(3)
Pump P-182 must be qualified SCII according to Position C.2 of Regulatory Guide 1.29 per NUREC-0800 as described in Section 3.2.
e G
4-10 m,
.,v.s v. i; &... a w 6% M
9, O
(4)
Pump P-182 must be capable of being manually loaded on either emergency diesel generator (EDG) feeding Bus Al or A2 under emergency conditions if both safety-grade AFW pumps fail to etart when there is a coincident loss of off-site power (FCE-to-NRC letter dated December 12, 1980 and NRC-to-PCE letter dated February 18, 1981).
(5) The not positive suction head (NSPil) required for F-182 at the design flow rate of 1,020 spa with a design head of 3,400 ft is 20 ft as determined by the pump supplier, Bingham-Willamette.
4.4.2 CONFICURATION AFW Pump P-182 is an eight-stage, horizontal, contrifugat pump with a double volute and a single suction first-etage impeller. A pump with eight stages is used to obtain the desired system flow rate and pressures with a'3,600 rpm motor.
the pump is driven by a 1,250-hp ac motor supplied from 4.16-kV Bus AS.
Pump design and operating data are provided in Table 9-4 Initial bearing lube! cation for F-182 is provided by' P-183, an installed motor-driven lube oil pump.
Power for P-183 is supplied from 480-V ac Bus B-30 (after RDC 86-004 is implemented, power will be supplied from B-43).
P-183 starts first to circulate pump lubricating oil when a start of P-182 is initiated.
P-182 starts when lube oil pressure is greater than 6 pais.
As P-182 comes up to speed, a shaft-driven lube oil pump begins to circulate F-182 lubricating oil.
When P-182 reaches normal operating speed and adequate oil pressure is attained by the shaft-driven oil pump (approximately 8 psis), P-183 automatically shuts off.
P-183 also automatically starts whenever P-182 is running and lube oil pressure drops to 3 psi.
A manyal override (438) on C-341 can be used to allow a local start of*P-182 if 4-11
>,.L.<:i, D :s m.. + siva al
P-183 fails to automatically start during a P-182 startup.
Cooling water flow to the pump bearing lube oil cooler is supplied from a portion of the discharge of P-182.
Controls and indications are located in the control room and on Panel C-3A1.
P-182 may be started from the control room or locally at 4.16-kV A5 switchgear.
- may also be stopped by using the emergency-stop push button on Panel C-341.
P-182 automatic trips occur on low suction pressure or low lube oil pressure.
The low suction pressure trip prevents pump operation with inadequa'a NPSH.
An override allows operators to manually testart l
P-182 at a lower flow rate if it has shut down on the low suction pressure trip.
The low lube oil pressure prevents pump and bearing damage due to low circulating lube oil pressure.
Under emergency conditions (both ESF AFW pumps have failed to start coincident with a loss of off-site power), power to start and operate P-182 can be supplied from 4.16-kV ESF Bus Al or A2.
An unloaded ESF bus supplied from an EDG is required to start P-182 because of the large power requirement and high starting current of the pump motor.
i Electrical loads may be added to the ESF bus after P-182 has started, but the total must be kept less than the maximum as specified in Trojan Off-Nornal Instruction ONI-55.
There are no automatic starting features asso:iated with this pump because of the large power requirement and high starting current of the pump motor.
4.5 MANUAL VALVES Manual valves in the Auxiliary Feedwater (AFW) System provide positive isolation for portions of the system to allow maintenance during plant operation.
e 4-12 as.. k.a.
ed o
4.5.1 DESIGN REQUIREMENTS (1) Manual valves in the flow path from the condensate storage tank (CST) to the tie-in point with the Main Feedwattar System must be locked open as required by Recommendation GS-2 of NRC-to-pCE letter dated October 3 1979.
(2) There must be no isolation valves in the common ATW pump suction piping from the CST.
(3) Manual valves that serve a safety-related function or could interfere with a safety function must be locked in the position required for automatic initiation.
4.5.2 CONFIGURATION RDC 79-064 provided limit switches for open or open/ closed position indicttion in the control room and the Remote Shutdown panel C-160, for manual valves in the flow path from the CST to the tie-in point with the Main reedwater System.
RDC 86-002 removed the common ATW pump suction isolation valve (ND-050).
Alarms are provided at C-160 for closed valven in the flow path from the CST to the Main Feedwater System.
Monthly inspections are performed to ensure all manual valves in the AFW System flow path are in their proper position per periodic Operating Test (p0T) 5-2.
4.6 AUXILTARY FEEDWATER FLOW CORTROL ISOLATION VALVES Each of the eight flow control valves. CV-3004 Al through D1 and A2 through D2, are required to automatically isolate a ruptured auxiliary feedwater (AFW) feed header to maintain the required flow to the intact steam generator feed headers.
The valves are also used to concrol AFW flow to ind!vidual steam generators.
4-13 L
- o-
' +, w - - d. ; e -e ao,
e.J
4.6.1 DESICW REQUIREMENTS (1)
The valves must be quallfled SCI in accordance with Position C.1.s of Regulatory Culde 1.29.
(2) The valver must meet the redundancy and single-failure criteria requirements discussed in Section 3.4.
(3) The palves must be safety-related and powered by separate Class 1E 125-V de power supplies in accordance with CDC 17.
(4)
The valves must be environmentally qualified for a harsh anvironment per pCE-1025 as described in Section 3.5.
(5)
The valves have remote status indication and are operable from the control room and panel C-160 in accordance with GDC 19.
(6)
The valves can be opened or closed against the differential pressure (2.170 psid) of the AFW pump running at shutoff head at the overspeed trip setpoint with the corresponding steam generacor depressurized.
4.6.2 CO?IFIGURATIO!!
The eight AFW flow control / isolation valves are 3-in., SCI slobe valves operated by Class is 125-V de motor actuators.
Con *.rols and open/close indications allow remote manual throttling for each valve from the control room or the Remote Shutdown panel, C-160.
CV-3004 Al through 01 are powered from 125-V de Train At CV-3004 A2 through D2 are powered from 125-V de Train 8.
During normal (standby) operation, all CV-3004 valves are full open.
High AFW flow rate in a braitch line, as sensed by flow switches FIS-3004 Al through D1 and FIS-3004 A2 through D2, isolates AFW flow 4-14
...u.u.is..& v..,
....roku A
.s.
through the branch line by closing the associated flow control /
isolation' valve.
The setpoint for the flow swicches is 500 spm.
If a valve in one train automatically closes upon high flow, interlocks prevent automatic closure of remaining CV-3004 valves in the same train.
A reset capability is provided to reopen automatically closed valves after the high flow fault has been cleared.
The reset circuitry is powered by Class 1E 120-V ac power supplies.
The valves are on the 59-f t level of the Maln Steam Support Structure and are qualified for operation in a harsh environment per pCE-1023 as described on Drawing E-2.
l 4.7 l
TURBINE STEAM SUPPLY ISOLATION VALVES
)
The Turbine Driver X-107A steam supply isolation valves are CV-1451 through CV-1454 and MO-3170.
These valves isolate steam supplied from the four main steam headers to Turbine Driver X-107A.
Control Valves CV-1451 through CV-1454 operate automatically to supply steam to this turbine driver.
MO-3170 is normally open and does not automatically operate.
1 4.7.1 DESIGN Rp0UIREMENTS l
4,7,1,1 CV-1451 throurh_CV-1454 (1)
Limit switches for control Valves CV-1451 through CV-1454, and Solenoid Valves SV-1451 through SV-1454 must be environmentally qualified per pCE-1025 as described in Section 3.5.
i (2)
CV-1451 through CV-1454. SV-1451 through JV-1454, and Accumulators T-166A through T-166D must be quellfled, SCI per Section 3.2, as required by Position C.1.s of Regulatory Guide 1.29.
4 i
4-15
~
.3s.
Js e nl.
S, GW...
o (3) CV-1451 through CV-1454 must operate independently of ac power as required by Recommendation CL-3 of NUREG-0611, as described in NRC-to-PGE letter dated October 23, 1980.
SV-1451 through SV-1454 de-energize on a loss of ac power and direct air to the bottom sides of the actuators of CV-1451 through CV-1454, allowing them to fail open on a loss of ac power.
(4)
SV-1451 through SV-1454. which operate CV-1451 through CV-1454. are powered from Class 1E 120-V ac power supply panel Y11 as required by CDC 17.
(5) CV-1451 through CV-1454 each have a dedicated accumulator (T-166A through T-1660) to ensure valve operation upon loss of instrument air. The accumulators are sized (15 gallons) to allow three valve motions (open-close-open).
(6) CV-1451 through CV-1454 must be shut during normal operation.
(7) CV-1451 through CV-1454 must meet piping pressure requirements per Section 3.1.
4.7.1.2 MO-3170 (1) Ho-3170 must be environmentally qualified per PCE-1025 as i
described in Section 3.5.
(2)
MO-3170 must be qualified SCI per Section 3.2 as required i
by Petition C 1.g of Regulatory Guide 1.29.
(3) MO-3170 must be required to be in the open position to ensure that X-107A and P-102A can be operated independently of ac power.
i 4-16
.,m a.. _
-.. m
I (4) The M0-3170 must be powered by a Class 1E 480-V ac power supply as required by CDC 17.
0 (5) The M0-3170 valve body pressure rating must meet piping l
pressure requirements per Section 3.1.
1 (6)
Redundant control switches and valve position indicators are l
located in the control room and on the Remote Shutdown panel (C-160) as required by CDC 13 and CDC 19.
l r
4.7.2 CONTICURATION L7.2.1 'CV-1451 Throuth CV-1454 l
Steam supply lines to X-107A from each of the four main steam headers are isolated by normally shut control valves.
These control valves.
CV-1451 through CV-1454, are 3-in.. air-operated gate valves that automatically open when K-107A receives a start signal.
Air is directed from the Instrument Air Systaa to the pistons of CV-1451 through CV-1454 by operation of Solenoid Valves SV-1451 through SV-1454 The solenoid valves receive power from 120-V ac panel Y11.
on a loss of power, SV-1451 through SV-1454 are automatically J
positioned to direct instrument air to the control valve actuators to l
r j
open CV-1451 through CV-1454
)
l A 15-ga11on accumulator in the air supply line to each control valve s
allows operation of the valve if instrummt air pressure is lost.
The sizing of ;he accumulators (T-166A through T-166D) allows three valve f
motions per control valve (open-close-open).
The accumulator size is based on a calculation performed as part of RDC 80-003.
Control switches for operation of SV-1451 through SV-1454 and, position I
l indication for CV-1451 through CV-1454 are located in the control room i
and at the Remote Shutdown Panel (C-160).
]
J j
4-17 s
. b sm, c a
. a 'c !
i The control valves are designed to open with a maximum differential pressure of 1.270 psid.
The maximum pressure ensures that the control valves can open with the maximum pressure in the steam generators and almospheric pressure downstream of the valves.
4.7.2.2 MO-3170 MO-3170 is the AFW Pump p-102A turbine driver stop valve.
It is a normally open. 4-in.. motor-operated globe valve. located downstream of Control Valves CV-1451 through C-1454 and upstream of MO-3071. the turbine trip and throttle valve.
The valve is in the open position to ensure that steam is supplied to X-107A to allow p-102A to operate independently of ac power.
MO-3170 is designed to open with a maximum differential pressure of 1.270 psid.
The maximum pressure across the valve ensures that it can open with the maximum pressure in the steam generators and atmospheric pressure conditions downstream of the valves.
A control switch (common for MO-3170 and MO-3071) and position indication light are located in the control room.
A separate control switch and position indication are located on panel C-160.
Electrical power for MO-3170 is supplied from 480-V ac ESP Bub B-23 (Class 1E power supply).
4,8 TURBINE TRIP AVD THROTTLE VALVE Turbino Trip and Throttle Valve MO-3071 is located in the steam supply line to X-107A downstream of MO-3170.
The valve automatically opens to provide steam flow to X-107A upon startup and rapidly closes under an overspeed condition to shut off steam flow.
4,8.1 DF0tCN_Rf0VIRE_MJUTO
~
(1)
MO-3071 must be qualified SCI per Section 3.2. as required by position C.1.g. of Regulatory Guide 1.29.
4-18
- 4. '.b a C">O
~: n
l o
(2) MO-3071 and the associated limit switches must be environmentally qualified per PCE-1025, as described in Section 3.5.
(3) MO-3071 is supplied with power from 125-V de Distribution Panel Dio, a Class 1E power supply a3 required by CDC 17.
A de power supply ensures that P-102A can operate independently l
of ac power for at least 2 he as required by Recommenda-tion CL-3 of NUREC-0611 described in WRC-to-PCE letter I
dated October 23, 1980.
f i
(4) Redundant control switches and valve position indicators for MO-3071 are located in the control room and at the Remote Shutdown Panel (C-160), as required by CDC 13 and CDC 19.
(5) Mo-3071 is shut during normal operation, and automatically opens on any P-102A start signal.
(6) MO-3071 cust meet the piping pressure requirements of Section 3.1.
4,8.2 CONFICttRATIOM M0-3071 is a 4-in., motor-operated globe valve that provides steam flow to X-107A upon a startup of P-102A.
The valve has a relatively i
slow stroke time (20 seconds) to provide control of steam flow until 1
),
the X-107A governor valve can adequately control steam flow.
It also automatically closes upon a X-107A overspeed condition to protect X-107A and P-102A from damage.
1 MO-3071 can open with a maximum differential pressure of 1,270 psid so the valve can open with maximum pressure in the steam generators and atmospheric pressure downstream of the valve.
i l
4-19 t
m e., _.i mm....... s.. ol '
MO-3071 automatically opens upon any of the automatic start signals listed for X-107A in Section 4.2.2.
To ensure it opens as required, the valve receives de power independently of ac power.
A control switch (common for MO-3071 and M0-3170) and position indication light are located in the control room. A separate control switch for MO-307! and valve position indication is located on Panel C-160.
4.9 SERVICE WATER SYSTEM TSOLATION VALVES There are four Service Water System isolation valves for the Auxiliary Feedwater System.
Two valves, MO-3045A and MO-30455, isolate the SCI water supply to the suction of engineered safety features AFW i
Pitmps P-102A and P-1028.
The remaining two valves MO-3060A and MO-3060B, f.solate service water cooling flow to the auxiliary coolers of P-102A/X-107A and P-102B/X-1078/I-173, respectively.
4.9.1 DESIGN REQUIREMENTS 4.9.1.1 Mo-3045A and M0-30458 (1) MO-3045A and MO-30458 must be qiialified SCI per Section 3.2, as required by Position C.1.s of Regulatory Guide 1.29.
(2) M0-3045A and MO-30458, and their associated limit switches, must be environmentally qualified per POE-1025, as described in Section 3.5.
(3)
MO-3045A and MO-30458 must be powered from class IE 480-V ac supplies as required by CDC 17.
(4) MO-3045A and MO-3045B must meet piping pressure requirements per Section 3.1.
4-20
.ma..si.
. & v n waw n a M k
(5)
The two valves. No-3045A in Train A and NO-30455 in Train 8 must meet redundancy and single-failure criteria described in Section 3.4.
4.9.1.2 MO-3060A and MO-30608 (1) MO-3060A and MO-30608 must be qualified SCI per Section 3.2 as required by Position C.1 3,of Regulatory Guide 1.29.
(2) M0-3060A and MO-30608 and the associated limit switch, must be environmentally qualified per PCE-1025. as describe 6 in section 3.5.
(3)
MO-30608 must be powered from a Class 1E 480-V ac power supply as required by CDC 17.
(4)
M0-30608 automatically opens upon any of the start signals associated with X-1078. as listed in Section 4.3.2.
(5)
MO-3060A and MO-30608 must meet piping pressure requirements per Section 3.1.
l 4.9.2 C0KFICURATION 4.9.2.1 MO-3045A and M0-30455 l
M0-3045A and M0-30458 are 6-in.
actor-operated gate valves.
They are normally shut to isolate the Cervice Water System supply to the suction of ESF ArW Pumps P-102A and P-1028.
Both valves have position l
indication and control switches in the control room.
MO-3045A l
receives electrical power from 480-Y ac EST Bus 8-25. and MO-30458 receives power from 480-V ac ESF Bus 8-26.
4-21
{
-.s. L e n + s e s u - - " a - !
The valves can open upon a maximum differential pressure of 100 psid; this capability is based on the Service Water System being in operation (both service water pump and service water booster, pump in service in one train) with one pump operating at the shutoff head, as verif 8 ed in Bechtel Cale.ulation 16-50.
These valves have no autamatic contcol functions associsted with their operation. They are only opened, from the control room, during emergency AFW System operation after the water level of the condensate storage tank (CST) has decreased to less than the low-low level alarm setpoint (<9 percent).
The operators then have a maximum of 30 min.
to open the valves from the time the low-low level al.asu is actuated before vortexing in the CST occurs.
The level setpoint was determined as part of pcE Cale'ulation TE-115 AFW Pump / CST Level Control Modift:4 tion for RDC 84-096.
A couaitment to the cLze requirement was made by PGE on the basis of the Safety Evalustion Report (SER) transmitted in the RRC-to PGE letter dated October 23, 1980.
4.9.2.2 MO-3060A and M0-3060B MO-3060A and MO-3060B are both motor-operated valves that isolate cooling water from the Service Water System to the auxiliary coolers of P-102A, X-107A and X-173.
MO-3060A is a 2-in, globe valves MO-30608 is a 6-in. butterfly valve.
The valves can open with a maximum dif ferential pressure of 100 psid: this capability is based on the Service Water Syctem being in operation (both service water pump and service water booster pump in service in one train) with one pump operating at the shutoff head, as verified in Bechtel calculation 16-50.
MO-3060A is electrically disabled and normally shut.
It is not required to be open since the lube oil coolers for P-102A and,X-107A are cooled by a portion of the P-102A discharge flow.
The automatic operating power was disabled by modifications performed under 4-22 1
- J D * * %* * *** * * * *
? # N y
MQ.*
0 o
RDC 78-020.
Because MO-3060A had been powered from an ac power source and was operated to allow cooling water from the Service Water System to the coolers for P-102A and K-107A, it was required to be disabled I
to meet NRC requirements of Recomnendation CL-3 of FUREC-0611, as l
described in FRC-to-PCE letter dated October 23, 1980.
This lineup allows operation of P-102A and K-107A to be independent of the ac power for at least 2 hr.
MO-30608 receives electrical power from 480-V ac EST Bus B-24.
It is normally shut when diesel-driven K-1078 is idle, to allow the jacket waterheaters to penheat K-107B.
MO-3060B eutomatically opens upon any of the K-107B automatic start signals listed in Section 4.3.2.
4.10 CONDENSATE STORACE TANK LEVEL INSTRUMENTAT!cN Redundant condensate storage tank (CST) level instrumentation is provided fort (1)
Engineered safety features auxiliary feedwater pump trips upon a low CST level.
(2)
Level indication in the control room and at the Remote Shutdown Panel.
(3)
Level alarms to alert the operators that a level limit is j
being approached.
4.10,1 DESICN_REQUIRFMUTTS I
(1)
Level Transmitters LT-5201 (Train A) and LT-5265 (Train B)
[
trip the associated ArW pump upon a low CST level or af ter failure of the CST.
(2)
The level transmitters and pump trip circuitry must be
{
qualified SCI; level indications and alaras must be SCII.
l I
4-23 i
.w h C b-44
(3)
Redundant level transmitters are located on opposite sides at the bottom of the CST, and are protected by enclosures.
(4)
Redundant CST level indication and low level alarms must be provided in the control room to allow operators to anticipate the need to make up water or to transfer to an alternate water supply.
The low level alarm setpoint should allow at least 20 minutes for operator action.
This requirement was promulgated in NRC to Put letter dated October 13, 1979, NRC Requirements for Auxiliary Feedwater Systam at Trojan Nuclear Plant, Additional Short-Term Recommendation No.1. and acknowledged in PCE to FRC letter dated December 31, 1979 Responses to October 3, 1979 ERC Questions Concerning the Trojan Auxiliary Feedwater System.
4.10.2 CONFICURATION RDC 84-096 upgraded the CGT level instrumentation from nonsafety grade / single channel to safety grade / redundant channel, to improve system reliability and to change the ArW pump trip from a low suction pressure to a low CST level.
The output of the transmitters indicates "low" if they break away from the CST.
The diesel ArW pump trips at 35 percent; the turbine ATW pump trips at the 30 percent level.
Override capability is provided to allow controlled operation below these levels.
A low-low level alarm at a 9 percent tank level allows the operator at least 30 minutes to switch to the backup Auction source before vortexing and loss of pump suction occurs.
The basis for the trip and alarm setpoints was determined as part of PCE Calculation TE-115, ATW Pump /C3T Level Control Modification for RDC 84-096.
e e
4-24 A-h-c>
- <L
l
-4.11 DIFFERENTIAL PRESSURE CONTROL Operation of the two ESF AW Pumps F-102A and P-1028 and electric A W i
Pump P-182, is controlled by separate differential pressure control circuits.
The circuits are similar since each ccapares the difference between steam generator pressure and the respective pump discharge pressure to a preset differential pressure to control the pump flow rate.
The design requirements listed in Section 4.11.1 concern only the ESF A N pumps.
Section 4.11.2 describes both the ISF AFW pumps and the F-182 differential pressure control circuits.
4 f
a 4.11.1 DESICN REQUIREMENTS i
1 (1)
Instrumentation for the differential pressure control of ESF AW Pumps P-102A and P-102B must be qualified 3CI per Section 3.2, as required by Position C.1.g of Regulatory l
Cuide 1.29.
1 l
t I
(2)
Instrumentation for the differential pressure control of E3F l
AW Punps P-102A and P-102B eust be powered from separate i
Class 1E 120-V ac EOF power sources, as required by CDC 17.
)
(3)
Instrumentation for the differential pressure control of E P AW Pumps P-102A and P-1025 must be environmentally qualified per PCE-1025 as described in Section 3.5.
l 1
i i
l (4) Two steam generator pressure inputs are available to each 4
E P AW pump pressure control circuit to meet redundancy and single-failure criteria requirements described in i
Coction 3.4.
I (5)
Redundant instrumentation for operation of ESF AW
]
Pumps P-102A and P-102B differential pressure control are I
)
t 4-25
. i. e e
+!
2.
located in the control room and at the Retote Shutdown Panel to moet the requirements of CDC 13 and'CDC 19.
4.11.2 CONFICURATION The differential pressure control circuits for ESF AFW Pumps P-102A and P-102B maintain a preset differential pressure (100 psid) between steam generator pressure and the pump dischars* pressure by controlling the speed of X-107A and X-1075, respectively.
The differential pressure cetpoint of 100 psid is based on ensuring that sufficient pressure is maintained in the steam generators to prevent leakage of reactor coolant through a U-tube (Westinghouse to Bechtel Letter POR-1710 dated April 25, 1973). The steam generator pressure signals to the P-102A differential pressure control circuit are supplied from Main Steam Headers A (PT-514) and C (PT-536). The steam generator pressure signals for the P-1025 differential pressure control circuit are supplied f rom Main Steam Headers A (Pf-516) and D (PT-545).
The two steam generator pressure inpat signals to the circuits are auctioneered to ensure that the purrp pressure control circuits receive the highest pressure signal. Auctioneering of the two steam generator input signals is required to prevent a low steam generator prest'are signal due to a circuit fault or a main steam line break f rom reducing AFV flow to the intact steam generators.
This auctioneering was installed as part of RDC 80-115.
The P-102A differential pressure control circuit receives power from 120-V ac Preferred Insttsment Bus Y11.
The F-1025 differential pressure control circuit receives power from 120-V ac preferred Instrument Bus Y22.
The P-182 differential pressure control circuit compares a preset I
differential pressure signal to the difference between cteam Generator C pressure and the P-182 pump dischars0 pressure.
An' output j
signal then varies the position of CV-2967 to malhtain the preset 4-26 i
- w... f.rh
.O d
i differential pres::"ce.
PDC-1967 IJ located in the control room for I
controlling the differenttei pressure of F-182.
4.12 VLOW INDICATION Flow instrumentation is provided for indlcation of MW System flow to i
each steam generator.
4.12.1 DrsIGN_RtoufREMENTS I
4 i
(1)
Redundant safety-grade AW System flow indication is required to monitor operation of the AFW System from the control room and the Remots Shutdown Panel (C-160) in accordance with NRC-to-PCE letters dated September 13. 1979 j
and October 3.1979 to implement NUREC-0578 recomraendations.
i l
(2) MV System flow instrumentation must be environmentally qualified and powered by Class 1E 120-V preferred instru: tent j
ac power supplies.
j (3) MV System flow instrumentation must be qualified SCI per Section 3.2. as required by Position C.1.g of Regulatory Guide 1.29.
4 1
4.12.2 CONFICURATION RDC 79-099 provided redundant SCI instrumentation to read combined flow from the A and 5 trains to each steam generator.
FT-3043A through 30430 provide indication at the Remote Shutdown Panel (C-160) and in the control room at Panel C-15.
FT-3043E through FT-3043H provide indication on the engineered safety features vertical bench board (Panel C-19).
Both channels share the same flow elements.
e 4-27 r nal.ms&Jn aa
- - 'k.w. ~ '
4.13 REMOTE SHUTDOWN STATION AFW System controls are provided in the control room.
In addition, controls are provided at the Remote Shutdown Panel (C-160) for use should the control room become inaccessible.
4.13.1 DESIGN REQUIREMENJT (1) CDC 19 requires that equipment be provided at appropriate locations outside the control room so that the plant may be shut down and cooled down.
Remote operation of the AFW System is required to bring the plan? to cold shutdown if the control room becomes inaccessible.
(2)
Panel C-160 must be powered from Class 1E power supplies.
(3)
Panel C-160 must meet 10 CFR 50, Appendix R requirements for separation and operability in the event of a fire.
4.13.2 CONFICURATION The Remote Shutdown Panel C-160 provides sufficient controls and indication to allow operation of the AFW System.
Lccal control for ATW Pump P-182 is not available at Panel C-160.
In addition Service Water System (SWS) Isolation Valves M-3045A and M-3045B cannot be operated from C-160.
They can be opened manually within prescribed time limits (30 min.) if required (refer to Section 4.9).
C-160 is in a mild environment and is thus exempt from 10 CFR 50.49 qualification (see PCE Drawing E-2).
Table 9-5 lists AFW System controls and instrumentation available at 1
the remote shutdown station, and their functions.
l l
l l
l 4-26 i
.. L usw. -:.
s a,...*
t i
The C-160 panel is prese-tly located in Room 89 (Fire Area T4) on the east side of the 45-ft 4evel of the Turbine Building, which is completely isolated from the Turbine Building general area.
According to the Scope Review of RDC 85-053, Appecilx R reviewer determined that the C-160 panel could become overheated in the event of a fire in the Turbine Building general area (Fire Area T1).
This overheating could possibly lead to the inoperability of both trains of the AW System.
RDC 85-052 will relocate panel C-160 to the 45-ft level of the Control Building across the hallway from the elevator in the old whole-body counting room to n.eet Appendix R separation and remote operability t
requirements.,
j
\\
4.14 AW PUMP P-182 DISCHARCE IS0!.ATION VAI.VES l
t t
i The two 6-in., motor-operated Cate Valves MO-2947A and M0-29478 fom a j
boundary between SCII (upstream) and SCI (valve and downstream)
)
portions of the AN System.
4.14.1 DESIGN REQUIREMENTS I
t t
(1)
MO-2947A and MO-29478 must be qualified SCI per Section 3.2, as required by position C.1.s of Regulatory Guide 1.29.
t i
I (2) MO-2947A and MC.29478 are nomally closed to isolate the SCII AW Pump P-182 discharge to the SCI trains.
(
4.14.2 CONFIGURATION I
The 7-182 discharge isolation valves provide isolation between SCI and t
j SCIA portions of the AW System.
They are provided with a Class 1E
[
480-V ac power supply for improved reliability.
Controls and
{
indications for operation of the valves are provided in the control Only one valve is opened when p-182 is used to supply feedwater room.
to the steam generators.
i i
[
4-29 i
>; f.- q u. -
'k!
' a-
4.15 ArW PUMP P-182 DIFFIPENTIAL PRESSURE CONTROL VALVE The dif ferential pressure control valve for AFW Pump P-182. CV-2967, operates to maintain a predetermined differential pressure, as compared to the actual differential pressure between Steam Cenerator C and the P-182 discharge pressure.
4.15.1 DESIGN REQUIREMENTS (1)
CV-2967 must be SCII per Position C.2 Regulatory Guide 1.29 as described in Section 3.2.
(2) CV-2967 must' meet piping Pressure requirements per Section 3.1.
4,15.2 CONTICURATION CV-2967 is a 6-in., air-operated globe valve installed in the discharge piping of P-182.
The differential pressure control of CV-2967 is described in Sectio'n 4.11.
Control room operators can take manual control of CV-2967. The valve may also be locally operated with a manual handwheel, if required.
t S
e G
4-30 i am,
. w.4. W
i l
t 5.0 SYSTEM OPERATION The OPERABILITY of the Auxiliary Feedwater (AFW) Systaa ensures that the Reactor Coolant System (RCS) can be cooled down to less than 350'F
(
from normal operating conditions in the event of a total loss of off-site power. The OPERASILITY of the condensate storage tank with I
the minimum usable water volume (196,000 gal.) and maximum tamperature i
(100*F) ensures that sufficient water is available to maintain the RCS t
under hot standby cond).tions for 2 he after a reactor trip, followed by a cooldown to 350*F in 4 he, with steam discharge to atmosphere, concurrent with a total loss of off-site power.
i 5.1 WORMAL OPERATICU i
I Under normal plant operating conditions, the AFW System is aligned to l
allow automatic initiation of flow to the steam generators.
The eight flow control valves are in the full open position.
Steam header isolation valves are anut. 1he steam stop valve is opet., and the turbine trip and throttle velve is shut.
The pump AVt0-MANUAL controllers are in the automatic position with a 100 paid setpoint.
The diesel-driven and turbine-driven AFW pump switches are in the 4
automatic position.
The maintenance lockout switches and low-CST level block switches on C-160 are in the NORMAL position.
The C-160 remote-local selector switches are in the REMOTE position.
5.2 NORMAL TRANSIENT OPERATICW The turbine-driven AFW pump is Train A.
Upon a START signal, the four steam supply yelves (CV-1451 through CV-1454) and the turbine. trip and throttle valve (MO-3071) open to supply steam to the Terry turbine.
5-1
.w. c +.
a..:
>M.-J
.r.
e The diesel-driven AFW pump in Train B.
Upon a START signal, the diesel starts and comes up to speed. The motor-operated Service Water System cooling water supply valve opens to provide flow to the diesel and pump lube oil coolers, the diesel engine jacket cooler, the turbo intercooler, and the speed increaser gear lube oil cooler.
4 When either Engineered Safety Features (gSF) AFW pump starts, the -
associated room supply and exhaust fans start to maintain room temperatures.
Normal supply of makeup is from the SCII condensate storage tank.
If the normal supply is lost, motor-operated valves are opened from the control room to provide service water as the SCI supply, i
A flow indication switch is located upstream of each motor-operated control valve (CV-3004-series valves) in the two supply lines to each stian generator.
A high flow in any line trips the FIS to close the respective motor-operated valve and lock the trips from the remaining 4
i valves in that train.
l During normal startup and shutdown, the electric AFW pump is used to supply feedwater from the CST to the steam generators to maintain i
desired steam eenerator levels or to place the steam generator in wet I
layup. The electric AFW pump is not qualified for design basis
)
events, and it functions to redace wear on the safety-grade pumps and drivers.
5,3 ABNORMAL AND EMERGENCY OPERATIONS The ESF AFW System is automatically initiated upon e y of the, following signals:
(1)
Safety injection (SI).
5-2
- i..;s a a2 a
e o
r (2)
Low-low level on two of three level transaltters in any I
(3)
Undervoltage on 4.16-kV Bus Al or A2.
(4)
Trip of both main feedwater pumps (unless bypassed).
Initiation of AFW isolates the blowdown and sampling lines for each steam Generator to prevent further water inventory leds.
P-182 can be manually loaded onto 4.16-kV EST Bus Al or A2 under emergency conditions.
l l
l l
l O
5-3
.-- s w C.a.444 m b -+e 4
i%M4 *:
s,. s.
i o
i 6.0 INSPgCTION AND TESTING This section summarizes design requirements for the inspection and testing of the AFW System, and describes those inspections and tests performed to verify proper operation of major components to ensure system compliance with the design basis.
6.1 DESIGN INSPTCTION AND TESTING REQUIREMENTS Inservice inspection of ASME Code Class 1. 2. and 3 components and inservice testing of ASME Code Class 1, 2. and 3 pumps and valves must be performed according to Section II of ASME Boller and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a(s), except where specific written relief has been granted by the NRC according to 10 CFR 50.55a(s)(6)(1).
The AFW Pumps P-102A and P-1025 and system valves (from Sections 4.6 4.7. 4.8. and 4.9) are currently tested according to PCE-1048. Inservice Testing Program for Pumps and Valves. Second Ten-Year Interval.
This program is based on requirements called out in Subsections IWP and IWV of Section KI of ASME Boiler and Pressure Vessel Code 1983 edition and addenda through the summer of 1983.
System piping is currently inspected according to PCE-1049. Inservice Inspection Program. Second Ten-Year Inte rval.
This program is also based on meeting the standards of ASME Code Section KI. 1983 edition and addenda through the summer of 1983.
The Trojan inservice inspection and testing programs contained in PCE-1048 and PCE-1049 are implemer,ted according to,PCE-8010. Nuclear Quality Assurance Program.
Compliance with the applicable requirement of ASME Code Section XI and appropriate addenda is ensured by an inspector as required by Paragraph IWA-2120.
e 6-1 r
- 1;.s'.S g dias %~S.
..:.9n.-. t M &
Qq;.cvs.k.
6.2 SYSTEN INSPECTION AND TESTING Inspection and testing of the AFW System and its ccaponents is performed according to PCE-1048 and PGE-1049.
The Technical Specifications specify the mode operability requirements, as described in Section 1.3.3, and the surveillance requirements for the AFW System and its components. The surveillance requirements list testing for the following AFW System components (1)
A7W Pumps P-102A and F-1025 (4.7.1.2.1).
(2)
Diesel Driver K-1078 (4.7.1.2.2).
(3) A7W Pump P-182 (4.7.1.2.3).
(4)
CST level requirements (4.7.1.4.1).
(5)
Service Water System requirements when supplying the ArW system
,(4.7.1.4.2).
The testing of the operability of the AFW System is performed according to the following Periodic Operating Tests (P0Ts):
(1)
POT-5-1, Pump and Valve Inservice Test.
(2)
POT-5-2, Valve Lineups and Inservice Testing.
(3)
POT-5-3 System Performance and Valve Inservice Test.
(4)
POT-25-2e, Switches $826 and 3841 - X-609 and K-633, trains A and B (SIS).
6-2
. s :.,-
.3. 1, i..,
- e rr. A.h e v. e * * *h :C+1hJ M
The tests conducted according to POT-5-1 satisfy the requiraments for the inservice testing of the following AFW components as defined by the Technical Specifications and PGE-1048:
(1)
Turbine-driven AFW pump and the associated steam supply check valves.
(2)
Diesel-driven AFW pump lube oil cooler water supply check valves, and associated suction check valves.
(3)
Electric AFW pump operability.
The tests conducted according to POT-5-2 satisfy the Technical Specifi-cations requirements verifying AFW System valve lineups and diesel fuel oil level for K-107B. and the inservice testing for time-cycle require-ments of the following valves as defined by PCE-1048:
(1)
CV-3004A1. Bl. C1. and D1.
(2)
CV-3004 A2. B2, C2. and D2.
(3)
MO-3045A and MO-3045B.
(4)
MO-3071 and MO-3170.
The tests conducted according to POT-5-3 satisfy requirements from both the Technical Specifications and PCE-1048.
One section of the POT i
satisfies the it-month survelliance requirement for operating the diesel-driven AFW pump as required by the Technical Specifications.
l Other sections of the POT satisfy the inservice testing requirements for l
the turbine-driven AFW pump steam supply check valves and both AFW Pumps P-102A and P-102B suction and discharge valves..t.dditidnal testing conducted per POT-5-3 verifies AFW pump design flow, proper position of the AFW pump discharge valves as indicated by position status lamps, and l
6-3
. >.. J a3. c y,.,:..
u - r.1
.w.cf
4 Proper automatic startup of AFW Pumps P-102A and F-1028 when a trip is inserted for both main feedwater pumps.
Calculations are then performed' from date during the previously described tests for pump suction conditions to ensure proper KPSH for the AFW pumps under all operating conditions.
The test conducted according to POT-25-2e verifles that Protection Relays K-609 and X-633 in the Safety Injection Systaa operate properly.
This test is a concern for the AFW System, since these relays automatically start AFW Pumps P-101A and P-1028 when a Safety Injection tignal is present.
These relays are tested to ensure that AFW Pump P-102A starts, that stoaa supply valves CV-1451 through CV-1454 open. and that AFW Pump P-1028 starts.
l I
l P
6-4 t
,_}
j. c,4 W I.. : r-C$tt. -
C i
J '
f i
i 7.0 DESIGN BASES EVOLUTION The following Trojan Nuclear Plant modifications affect the configuration of the Auxiliary Feedwater System.
A summary of each, with an explanation of its effect on the system design bases (if any),
is provided.
h 7,1 COMPLETED AND CLOSED OUT Rjg),
RDC 75-191 SVMMARY:
Modified the A W flow instrument wiring to prevent a ground fault from disabling the indication.
E7FECT OF BASES:
None.
RDC 75-276
SUMMARY
- Added inteslocks to service water valves for AFW i
turbine and diesel drivers.
l E[FECT ON BASES:
None.
[
l RDC 76-124 SUMKARY:
Replaced AFW turbine wheel due to wear damtse and high backpressure preblems.
i EFFECT ON BASES:
None.
i f
RDC 76-164
SUMMARY
Modified the AFW diesel exhaust line from 8 in.
to 12 in, to reduce crankcase backpressure.
i h
i I
f EFFECT ON BASES:
None.
I I
l I
ROC 76-175 SUKKARY:
Provided maintenance lockout switches for the l
l ESF AW pumps.
[
i i
EFFECT ON BASES:
None.
l l
7-1 M
- Wis~'-
h 48 -
o e
I RDC 76-210
SUMMARY
11odified valve lineup for the Terry turbine drain line.
1 EFFECT ON BASES:
None.
RDC 76-217
SUMMARY
Added oil reservoir to the Terry turbine governor actuator.
i Ef,FECT ON BASES:
None.
RDC 76-225 $UNMAM:
Relocated the govervior cabinet for the turbine-driven AIV pump behind Panel C-160 for temperature considert.tions.
M OW BASES:
None.
RDC 76-226
SUMMARY
Added interlocks for service water valves for f
AFV turbine and diesel.
EFFECT ON BASES Wone.
RDC 76-227 SUto WIYt Installed jacket water heaters on K-107F co improve, quick startup capability by maintain!;;g jacket water warm when K-1075 is idle.
EFi'ECT OW BASES:
None.
RDC 76-241
SUMMARY
Reinstalled original turbine Alv pump ramp generator / signal converter.
EFFECT ON BASES:
None.
e 7-2
..xdawe.
M-Y
.- - ~
o RDC 76-261
SUMMARY
Replaced AFW turbine trip and throttle valve relay enclosure with a watertight, corsesion-resistant enclosure.
EFFECT ON BASES: None.
I RDC 76-265
SUMMARY
Added drain coolers to X-107A steam line drains to reduce humidity in F-101A pump room.
Coolers are supplied cooling water from Service Watar System.
Efffyf ON BASES Nane.
l RDC 76-277
SUMMARY
Added ramp generator'to K-1078 speed control i
ci cultry to prevent overspeed during diesel startup.
l EFFECT ON BASES: None.
I P
RDC 76-278 SUMKARY:
Added a soakback pump for Diesel Driver K-107B to improve quick startup capability by tirculating lubricating oil when K-1078 is idle.
i EFFECT ON BA,111t None.
{
t RDC 76-285
SUMMARY
Raised elevation of the diesel AFW pump speed increaser auxiliary oil pump to prevent flooding c! the motor.
F i
EFFECT ON BASES:
None.
1 1
RDC 76-293
SUMMARY
Redesigned X-1075 diesel annunciator circuitry l
to indicate when diesel "NOT READY FOR AUT0 START".
[
EFFECT ON BASES:
None.
t 7-3 n.
. eA %E:
-# + L
.c s.
RDC 76-345
SUMMARY
Provided cutout for AFW autostart on loss of both main foedwater pumps, without activating maintenance lockout.
EFFECT ON BASES:
None.
RDC 76-515 SVKMARY:
Provided cutout for ATW autostart on loss of both main feedwater pumps, without activating maintenance lockout.
EFFECT ON BASEST None.
RDC 77-139
SUMMARY
Major modification installed electric motor-driven ATW Pump P-182, installation includedt P-182 suction piping from CST.
P-182 discharge piping to the diesel-driven ATW Pump P-1028 discharge piping.
- ATW Pump P-182, associated valves, and 1,250-hp ac motor.
Additional cubicle (A510) and circuit breaker in the 4.16-kV Bus AS rwitchgear.
- Local instrumentation, transmitters, and associated sensing linest instrument air supply lines for P-182 discharge control valva, CV-2967.
- Auxiliary lube oil Pump P-183 and associated pressure switches.
- Local temperature monitoring Panel C-341, for P-182 pump and motor.
EFFECT ON BASES:
Zeplementation established design baseline for AFV Pu=p P-182.
No effect on system design bases.
7-4
. /.. :.:C2 J
i i
RDC 77-162 BVMMARY:
Installed manometer to monitor AFW diesel i
l crankcase pressure.
EFFECT ON BASES:
None.
i RDC 78-020 SVMMARY:
Installed piping and associated equipment to turbine-driven AFW Pump P-102A to allow. cooling of F-102A and K-107A lube oil coolers from P-102A discharge flow.
Also removed electrical automatic actuation of NO-3040A, isolating Service Water System cooling water flow fron P-102A and K-107A lube oil coolers.
i EFFECT ON BASEST This modification enhanced the redundancy design basis for the AFW system, and specifically the electrical independence recommendations of BTP Ass-10 (see Section 3.4), by removing the ac-powered service water booster pumps as a required source of lube oil cooling for operation of P-102A and K-107A.
RDC 79-032 SVMMARY:
Added pipe supports for AFW diesel breather tube assembly.
EFFECT CW BASES:
None.
RDC 79-043 SVMMARY:
Installed flex base on AFW diesel fuel return line to avoid vibration cracking.
EFFECT OW BASES:
None.
KDC 79-063
SUMMARY
Added low suction pressure trips for the ESF AFW Pumps.
EFFECT ON BASES:
None.
7-5
- o m
..v.
ww - +
j l
RDC 79-064 3UMMARY:
Added limit switches for manual valves in the f1'ow paths of the two E8F AW pumps from the CST to the inlet piping of the Main Feedwattr System.
Installed I
indications in the control room and at Panel C-160.
I EFFECT OW BASES:
None.
4 RDC 79-099 StnogRY:
Upgraded power supplies for A W flow inattunen-i t
tation to Class 15.
Installed redundant, safety-grade l
j finw indicators in the control room to allow monitoring of i
AW System flow rate to all four steam generators.
1 j
EFFECT ON BASES:
Implemented redundancy requirements (see
(
Section 4.12).
j r.DC 80-003
SUMMARY
Provided for 2-hr operation of A N Pump P-102A i
independent of ac power.
Included e
i
- T.eplacement of motor operators for the X-107A steam inlet valves with air operators (CV-1451 through e
t CV-1454).
l 1
f
. Installation of solenoid valves sv-1451 through l
SV-1454 Air Accumulators T-166A through T-166D, and I
associated piping, tubing, and supports for the air
[
]
supply to CV-1451 through CV-1454.
EFFECT ON BASES:
Established design basis for the 2-hr operation of the turbine-driven AW Pump P-102 independent of ac power.
?
I RDC 80-024 2]LMMARY:
Changed power supply for AW Pump P-102A L
discharge pressure transmitter PT-3083A from 120-V Instrument ac Bus Y02 to 120-V Preferred Instrvrent ac j
Bus Y11.
l l
1 EFFECT cv Basts:
None.
l 1-t i
7-6 5
h
..6. o'
>-x
=
RDC 80-054 $UMMARY:
Installed a guard pipe around AFW Pump P-102A discharge piping located in the F-1023 paw room.
EFFECT OW BASES:
Thic modification Laplaaented BTP ASB-10 redundancy recomeendations with regard to a high-energy line break of the P-102A discharge piping.
It also l
t implemented PGE-1004 analyses for pipe whip and det impingement.
l-RDC 80-061
SUMMARY
Installed a second discharge piping connection for AFW Pump P-182 to allow discharge to P-102A discharge piping (as well as to P-1023 discharge piping).
Also installed check valves and motor-operated gate valves (MO-2947A and MO-29478) for isolation of P-182 discharge headers from the RSF AFW trains.
ETTECT ON BASES:
hone.
I i
a RDC 80-086
SUMMARY
Installed new battery and battery charger for Diesel Driver K-1078 starting system.
EFFECT ON BASEST None.
Installed to improve reliability.
RDC 80-115
SUMMARY
Provided a redundant steam generator pressure l
signal input to each ESF AFW pump differential pressure p
control circuit.
Also, entailed auctioneering circuitry 1
1 to account for spurious low steam generator pressure s'gnnis or main stema line breaks.
r i
EFFECT ON BASES:
None.
Installed to enhance system reliability.
i i
l 7-7 i
l t
RDC 80-104 SUMMAR1: Modified loss of control power annunciator to actuate "NOT READY FOR START" for AFW diesel and turbine.
t EFFECT ON 8ASES:
None.
i RDC 81-070
SUMMARY
- Added new power feeders for FIS-3004 Al through 01 and A2 through D2.
i EFFdCT ON 8ASES:
None.
RDC 81-073
SUMMARY
Replaced AFW flow transmitters with
[
environmentally qualified units.
EFFECT ON BASES:
Implamented environmental qualification requirements.
1 RDC 81-090
SUMMARY
- Modtfied turbine AFW pump tubing to eliminate governor oil overflow problems.
L EFFECT ON BASES:
None.
l RDC 82-016 gytggp.It Nodified AFW pump p-182 circuitry to provide I
smergency stop at C-341; installed new luba oil pump start /stop switch at C-341.
t
,EFFECT ON BASES:
None.
RDC 82-059
SUMMARY
Provided pipe support modifications to AFW f
instrument lines.
j EFFECT ON BASES:
None.
e I
7-8 su
. n+
.i
1 i
RDC 82-062 SVMMA]tY:
Modified A N diesel alarm circuitry to annunciate on loss of soakback pump.
EFFECT 08.BAlts:
None.
RDC 82-063 Mt Nodified AFW turbine alarm circuitry to indicate I
turbine "NOT READY FOR START".
l EFFECT ON BASEST Wone.
RDC 'J3-014
SUMMARY
Modified plyw supports for AFW diesel cooling jacket drain, i
RFFECT OW BASES:
- None, f
RDC 83-025
SUMMARY
Modified control circuitry for M0-3071 M0-3170.
EFFECT OW BASES:
None.
t RDC 83-046
SUMMARY
Added an enclosure for exposed AFW piping which was not in a vital area enclosure.
[
EFFECT ON BASES:
This modification was implemented for safeguard reasons.
i RDC 84-100
SUMMARY
Reloceted selected redundant circuit raceways, l
1 l
EFFECT ON BASES:
This modification implemented Appendix R review design requirements.
f f
RDC 84-104
SUMMARY
Replaced eight CV-3004 valves.
l l
\\
I EFFECT ON BASES:
Implemented environmental qualification requirements.
i I
t I
f 7-9 i
..I
- f.. w..
u n,
RDC 84-105
SUMMARY
Replaced eight AFW flow differential pressure switches for the CV-3004 valves.
EFFECT OW BASES:
Implemented environmental qualification requirements.
RDC 84-108 CUMMARY:
Installed seals on AFW steam inlet valve limit switches and AFV flow transmitters.
EFFECT OW BASES:
Irplemented environmental qualification requirements.
o 7,2 PDCs TNSTALLED BUT NOT CLOSED OUT RDC 84-096 EqFMARY:
Installed redundant Level Transmitters LT-5201 and LT-5265 on the CST, powered from 120-V preferred instrument ae.
Replaced low suction pressure trips for the ESF AFW pumps with low CST level trips for each ESF pump.
KFFECT ON BASES:
Wone.
Installed to enhance system reliability.
RDC 06-002
SUMMARY
Removed the common ESF AW Pump Isolation Valve KD-050 and replaced it with an orpansion joint.
FFFECT ON BASES:
This modification implemented the design requirement for no manual isolation valves in the common suction piping from the CST to the ESF AFW pumps (see Section 4.5.1) to preclude common mode failure of AFW pump flow.
It also implemented an upgrade of the piping from the expansion joint at the CST to the AFW pumps tp SCI.
9 7-10
[
i i
l
\\
l l
i l
RDC 46-024
SUMMARY
Added a pressure barrier to the diesei AFW pump room.
t l
EFFECT oW PA3nt Mone.
Implemented to maintain diesel i
t AFW pump room as a alld SQ environment.
7.3 RDCs INCOMPLETE OR PENDING i
i l
RDC 84-044 8UMMARY:
Changes scale on MW flow instrumentation to i
provide linear indication.
{
i l
EFFECT ON Basts:
None.
t i
i
[
3DC 84-091
SUMMARY
Installs decouple switch for F-102A.
Modifies I
circuit design to prevent fire-induced circuit damage.
I EFFECT ON BASES:
Implements Appendix R requirements to I
allow circuit operation without fuse replacement.
f RDC 85-029
SUMMARY
Provides ATW3 mitigation system actuation circuitry (AMSAC).
I t
IFFECT OW Bast 3:
Not yet evaluated.
t i
i RDC 85-051
SUMMARY
Changes power supply to AfW flow indication l
}
sensing lines heat tracing.
j f
j EFFECT ON Basts:
None.
t l
1 7-11
- - b - F'-#
^ " ' "-
~
l t
1 i
t RDC 85-052 SVMMARY:
Relocates Panel C-160 to the 45-f t level of the Control Building.
Installs additional instrumentation at C-160 to upgrade it to a remote shutdown station per i
f Appendix R requirements.
EFFECT ON BASES:
When implemented, will redefine design 7
baseline for remote shutdown requirements (see Section 4.13).
i RDC 86-003
SUMMARY
Implements various AW reliability improvements, i
including:
[
1
- Bypass of diesel engine jacket water high temperature c
trip on autostart (P-1025).
- Resetting thermal overload relay for MO-3060B to I
maximum current rating allowed by Appendix R l
requirements (P-1025).
i
)
- Removal of one timing relay in P-1028 trip circuitry.
I i
)
- Bypass of thermal overload relays for M0-3071 (P-102A).
- Sealing of Transmitters PT-3083A, PT-3072A, and ATB-701 f
j in P-102A pump room f rom high-moisture environment.
s l
- Common improvements to replace the AFW pump suction and
[
j discharge transmitters.
ETTECT CN PA$tSt None.
l 4
1 j
RDC 86-004 gjap*fL:
Modifles AW Pump P-182 to provide conunon power l
supply for Auxiliary Lube 011 Pump P-183, local indication j
of suction pressure, local override of the low suctior.
j j
pr.ssur. trip, and local indication for Csf lev.1.
t i
f
((FECT ON BASES:
None.
i 1
l t
i t
i i
7-12 l
A
.a
. g.
mo
o i
i RDC 86-036 E[tgA31:
Installs DC monitor relays for the governor i
control power circuitry for both K-107A and K-1075.
i Provides indication to operators if any fuses have blown in the governor control power circuits causing machines to slow down.
l KFFECT ON 8ASES:
Wone.
1 RDC 46-039 StAINARY:
Replaces AFW turbine steam supply check valves with newer, upgraded-type check valves.
6 EFFECT ON BASES:
None.
l t
RDC 86-043
SUMMARY
Modifies control circuitry for MO-3071 and I
NO-3170.
i
(
EFFECT ON SASES:
None.
l RDC 86-047
SUMMARY
Installs de monitor relay coil for X-107A I
electronic speed monitor.
I 1
EFFECT ON BASES:
None.
I l
4 l
J d
l 6
}
7-13 1
J
(
u+A A
e o
N lll1 p'q sp K
A ril B
Elill ll 11 t!!eg
[
3 n
t in J:i i wN 4 t, gg- >
t
[p
{
4 er 7, 4,3 6 l 3 Q31 m
5-) k,(e i
y, i
~7:
I
!. i isli x
\\ll 5
i g
18 a
i i
m
!y
'l P,
j l
c SX st
- s.,
s t
}y
, u.%
al j h
~
Il p
u s.
ett la x-4
=Il r /.
e 133 8~+1 55s b
s,
$1ejg
~
- + > - =
g 4
,1
^+5 q r-,
+ L I
i,
i 3
(;.
~
~
g
@ Z m
,4 8't*5 I!
bwg:
g*
l e,:
+
l I.
ga
~
b
--07 ;
{
y " f
, ~8 M g
n il a
- L n
B E
s s
a 4
I
-a a
r N
g_-
gg I
~
M+
=
m s.-t--5 h
M g
n 2
I j
i l
s
-.y W
l P
Uma 0
e9
l L
l i
TABLE 9-1 AUXILIARY FagDWAfgA PUMP 3 P-101A AND P-1025 4
DATA SHggT
[
i
{
Type Morizontal centrifugal i
Flow rate (design) 960 spa (including to spa recirculation) r 3
Pump casing pressure (design) 2,000 pelg
[
Rated head at, design flow rate 3,400 ft i
I i
Shutoff head 3.910 ft h
WPSH required at design flow rate 25 ft t
l Minimum rpsH available 26 ft l
Suction temperature (normal) 70'r t
l Speed 4,560 rya l
l
]
Discharge pressure at shutoff head 1,700 psig
)
- l Lube oil Circulation 3.2 spa i
l Temperature at bearing outlet 150'F 4
Heat transferred to oit 12,000 stu/hr i
Cooling water flew to lube oil ceoler 5 spa l
)
MaxLaun cooling water inlet temperature 85'r i
Reservoir essacity 17.7 gal.
l l
1 i
t t
t I
F l
9-1 L
l l
l l
TASLE 9-2 TURBINE DRIVER X-107A i
+
Type Single-stage, i
noncondensing Steam inlet pressure f
Design 125 to 1.320 psia l
Operating (normal) 900 to 1.100 psia i
Nazimum steam inlet temperature 590'r f
Backpressure on casing I
Maximum 165 peig Wormal 0 peig
(
I I
Steam moisture condition (maximum) 1/2%
1 t
Rated bhp (at 1.305 pois and 58C*r 1,045 f
steam inlet temperature) j Rated speed 4.560 rpa Speed operating range 2.660-4.560 rpm j
i Electrical trip speed 5.100 rya
{
Mechanical trip rpeed 5.500 rya
[
l t
i I
1 r
I i
l I
1 I
I s-2 x.. e,
(
1
O i
i l
TABLE 9-3 i
DIESEL DRIVER K-1078 DATA SHEET
}
Rated bhp 1,579 i
Rated speed 1,200 rpm l
Operating speed range 450 to 1.200 rpm Diesel engine overspeed trip 1.350 rpm l
Diesel fuel type No. 2 diesel fuel r
I Diesel fuel oil day tank volume 500 sailons Normal oil pressure 45 t,5 psis l
I Servlee Water System Flow Rates to X-1078 (Des 1Lg) l Diesel engine jacket cooler 230 spa l
Speed increaser lube oil cooler 15 gym
[
r l
Diesel engine lube oil cooler and intercooler 67 spa
{
r I
s,..d iner.as.e r
Ratio (rya) 1 3.8 Lube oil flow 16.75 sps I
Lybe oil cooler tenweratures Cooling water flow (inlet / outlet) 90'F/94*F j
l Lube oil flow (inlet / outlet) 150'F/132.S'T f
Flow orifice l
Differential pressure (maximum) 1,470 psid j
Flow (maximun) 150 spa l
9-3 1
l
..I s.
n
o TABLE 9-4 AFW PUMP P-182 DATA SHEET Flow rate (design) 1,020 spm (including 140 spm recirculation)
Speed 3,560 rpm Head at design flow rate 3.400 ft WPSH required at design flow rate 20 ft NPSH available 25 ft Pump casing pressure (design) 2,000 psig Suction temperature (design) 90*F Cooling watne flow to lube oil cooler 10 spm Maximum cooling water inlet temperature 90*F Driver Type Electric motor, 1,250 hp Hotor cervice f astor 1.15 Speed 3,560 rpr.
Eunp_ trips 4
Lcw oil peessure
<3 psis Low suction pressure
<12 psia i
l 4
e 9-4 I: s t. w
< n n, u
TABLE 9-5 AUKILIARY FEEDWATER SYSTEM INSTRUMENTATION AT REMOTE SHUTDOWN STATION (C-160)
Instrument Function Train A Controls
- Manual / auto control for P-102A
- Manual / auto control for CV-1451 through CV-1454
- Manual control of CV-3004A1, Bl. C1, D1
- Manual differential pressure control for P-102A Indication
- Position indication for valves operated from RSS
- Position indication for manual AFWS valves
- P-102A suction and discharge pressure
- X-107A steam supply and exhaust pressure
- P-102A/SG differential pressure
- K-107A operating speed
- CST level
- SG pressure
- SG 1evel
- ArW flow to SGs Train B Controls
- Manual / auto control for P-102B
- Manual control for CV-3004A2, B2, C2 D2 i
- Manual differential pressure control for P-102D Indication
- Position indication for valves (perated from RSS
- Position indication for manuel AFWS valver
- P-102B suction cnd discharge pret'sure
- P-102B/SG differential pressure
- K-107B operating speed
- X-1078 diesel fuel oil day tank le%91
- CST level
- SG pressure r
- SG level
- APW flow to SGs e
9-5
...a.- 3
e
10.0 REFERENCES
10.1 CODES. STANDARDS. AND REGULATORY REFERENCES 10.1.1 CODES AND STANDARDS 10.1.1.1 ASME Boiler and Pressure Vessel (B&PV) Code,Section VIII. Division I, 1968 and Addenda through 197*
10.1.1.2 ASME B&PV Code,Section III Class 2 Nuclose Power Plant Components, Containment Penetrations, 1971 and Addenda through summer of 1971.
10.1.1.3 ASME B&PV Code Section XI 1977 Edition, IWA-7210, and 1983 edition and Addenda through summer of 1983.
10.1.1.4 Draf t ASME Code for the Inservice Testing of Pumps in Nuclear Plants, April 1970.
10.1.1.5 Draf t ASME Code for the Inservice Testing of Valves in Nuclear Plants. June 1970.
4 10.1.1.6 ANSI B31.1.0, Power Piping Code, 1973.
10.1.1.7 ANSI N45.2.11-1974, Quality Assurance Requirements for the Design of Nuclear Power Plants.
10.1.1.8 IEEE 279-1971 Criteria for Peotection Systems for Nuclear Power Generating Stations.
10.1.1.9 IEEE 308-1971, Standerd Criteria for Class 1E Power Systems for Nuclear Power Generating Stations.
10-1
~;;
m.
i
o 10.1.1.10 IEEE 323-1971 Trial Use Standard for Qualifying Class IE Electrical Equipment for Nuclear Pover Generating Stations.
i 10.1.1.11 IEEE 323-1974 Standard for Qualifying Class 1E Electrical Equipment for Nuclear Power Generating Stations.
10.1.1.12 IEEE 338-1977, Standards for the Periodic Testing of Nuclear Power Generating Station Safety Systems.
10.1.1.13 IEEE 344-1971 Guide for Seismic Qualification of ClasJ 1E Electrical Equipment for Nuclear Power Generating Stations.
10.1.1.14 IEEE 344-1975 Recommended Practices for Seismic Qualification of Class 1E Electrical Equipment for Nuclear Power Generating Stations.
10.1.1.15 NEMA Standard SM-22-1970 Single-Stage Steam Turbine for Mechanical Drive Servios.
10.1.2 REGULATORY REFERE CE_S, S
10.1.2.1 Code of Federal Regulationc. Title 10 (10 CFR).
10.1.7. 1.1 10 CFR 50.48 Fire Protection.
10.1.2.1.2 10 CFR 50.49. Environmental Qualificat.8cn of Electric Equipment Irportant to Safety for Nualete Power Plants.
10.1.2.1.3 10 CFR 50.55a. Codes and Standards.
10-2
~
'e
10.1.2.144 10 CPR 50, Appendix A, General Design Criteria for Nuclear Power Plants.
10.1.2.1.5 10 CFR 50, Appendix B Quality Assurance Criteria for Nuclear Power Plant and Fuel Reprocessing Plants.
10.1.2.1.6 10 CFR 50 Appendix R, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, Sections III.C, III.J. and III.L.
10.1.2.1.7 10 CFR 100, Reactor Site Criterit.
10.1.2.2 USERC Regulatory References.
10.1.2.2.1 USNRC Regulatory Guide 1.29, Seismic Design Qualificacion (Rev. 3), September 1978.
10.1.2.2.2 USERC Regulatory Guide 1.30, Quality Assurance Requirements of the Installation, Inspection, and Testing of Instrumentation and Electrical gquipment, August 1972.
10.1.2.2.3 USKRC Regulatory Guide 1.62, Manus 1 Tnitiation of Protective Actions October 1973.
10.1.2.2.4 UcNic kogulatory Guide 1.75, Physical Independence of Electric Systems (Rev. 2), September 1916.
10.1.2.2.5 USMRC Regulatery Cuide 1.89, Qualification of Clads 1E Equipment for Nuclear Power Plants November 1974, 10.1.2.2.5 USNRC Hegulatory Guide 1.93, Availability of glectric Power Sources, December 1974.
e I
10-3 a
u.
,.. r
10.1.2.2.7 USNRC Regulatory Guide 1.97. Instrumentation for Light-Water-cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident (Rev. 3), May 1983.
10.1.2.2.8 USNRC Regulatory Guide 1.100. Seismic Qualification of Electrical Equipment for Nuclear Power Plants (Rev. 1),
August 1977.
10.1.2.2.9 USNRC Regulatory Guide 1.106. Thermal Overload Protect!on i
for Electric Motors on Motor-operated Valves (Rev. 1).
March 1977.
10.1.2.2.10 USURC Regulatory Guide 1.117. Tornado Design Classification (Rev. 1), April 1978, 10.1.2.2.11 NUREG-0578. THI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations, Section 2.1.7 July 1979.
10.1.2.2.12 NUREG-0588, (Category 1), Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment.
10.1.2.2.13 NUREG-0611 Geraric Evaluation of Fesdwater Transients and Small Dronk Losa-of-Coolant Accidents in Westinghouse-Designad Operatiat Plants, January 1980.
10.1.2.2.14 NUREG-0737 Clarification of TMI Action Plan Requirements,Section II.E.1.2 November 1980, 10.1.2.2.15 NUREG-0800, Stkndard Review Plan Section 10.4.9, Auxiliary Feedweter System (PWR), (Rev. 2), July 1981.
10-4
)
w a
10.1.2.2.16 USERC Branch Technical Position (BTP) APCSB 10.5-1 Appendix A. "Guidelines for Fire Protection for Nuclear Power Plants Docketed Prior to July 1,1976", A Jgust 23, 1976.
10.1.2.2.17 USERC Branch Technical Position (BTP) ASB 10-1, "Design Guidelines for Auxiliary Feedwater System Pump Drive and Power Supply Diversity for Pressurized Water Reactor Plants".
10.1.2.2.18 USNRC Division of Operating Reactors (DOR) "Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors", 1979.
10.1.2.2.19 USNRC IE Bulletin No. 79-02, Pipe Support Base Plate Designs Using Concrete Expansion Anchor Bolts March 8, 1979.
10.1.2.2.20 USNRC IE Bulletin No 79-04, Incorrect Weights for Swing Check Valves Manufactured by Velan Engineering Corporation, March 30, 1979.
10.1.2.2.21 USNRC IE Bulletin No. 79-07, Seismic Stress Analysis of Safety-Related Piping, April 14, 1979.
10.1.2.2.22 USNRC IS Bulletin No. 79-14, Seismic Analyses for l
As-Built Safety-Reisted Piping Systems July 2, 1979.
10.1.2.2.23 USNRC IE Bulletin No. 80-11. Masonry Wall Design, May 8, 1980.
10.1.2.2.24 USNRC IE Information Notice No. 80-21, Anchorage,and Support of Safety-Related Electrical Equipment. -
10-5 s
. aes 4.*.
M
10.1.2.2.25 USNRC IE Information Notice No. 84-54. Deficiencies in Design Base Documentation and Calculations Supporting Nutlear Power Plant Design.
10.2 P(.E INTERNAL DOCUMENTS AND TECHNICAL MANUALS 10.2.1 PCE/ TROJAN REPORTS i
10.2.1.1 Trojan Updated Final Safety Analysis Report (UFSAR).
t Docket No. 50-344 10.2.1.2 PGE-1004. Trojan Nuclear Plant Analyses of Pipe System Breaks. Outside Containment.
10.2.1.3 FGE-1012. Trojan Nuclear Plant Fire Protection Plan.
10.2.1.4 PGE-1022. Inservice Testing Program for Pumps and Valves.
10.2.1.5 PGE-1025. Trojan Nuclear Plant Environmental Qualification Program Hanual.
10.2.1.6 PGE-1028, Regulatory Guide In House Positions.
10.2.1.6.1 IHP 1.28-2-1 Regulatory Guide 1.28. Quality Assurance Program (Design Lnd Construction). December 31. 1981.
10.2.1.6.2 IHP 1.29-3-1. Regulatory Guide 1.29. Seismic' Design Qualification. September 1.1982.
10.2.1.6.3 IHP 1.30-0-2 Regulatory Guide 1.30. Quality Assurance Requirements for the Installation. Inspection and Testing of Instrumentation and Electrical Equipment. Sep,tember 1 1982.
10-6
-.i
., a
a
=
10.2.1.6.4 IMP 1.32-2-1, Regulatory Guide 1.32, Criteria for S'afety-Related Electrical Power Systems for Nuclear Power Plants, September 1, 1982.
10.2.1.6.5 IMP 1.59-2-1, Regulatory Guide 1.59, Design Basis Floods for Nuclear Power Plants September 1, 1982.
10.2.1.6.6 IRP 1.60-1-1, Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants, September 1, 1982.
10.2.1.6.7 IMP 1.62-0-1, Regulatory Guide 1.62 Manual Initiation of Protective Actions September 1, 1982.
10.2.1.6.8 IHP 1.64-2-2, Regulatory Guide 1.64, Quality Assurance s
Requirements for the Design of Nuclear Power Plants, September 1, 1982.
10.2.1.6.9 IHP 1.75-2-2, Regulatory Guide 1.75, Physical Independence of Electric Systems, December 31, 1984.
10.2.1.6.10 IMP 1.76-0-1, Regulatory Guide 1.76, Design Basis Tornado for Nuclear Power Plants, September 1,1982.
10.2.1.6.11 IMP 1.89-1-1, Regulatory Guide 1.89, Qualification of Class 1E Equipment for Nuclear Power Plants March 1986.
l 10.2.1.6.12 IRP 1.93-0-1, Regulatory Guide 1.93. Availability of Electric Power Sources. September 1,1982.
10.2.1.6.13 IKP 1.97-3-1, Regulatory Guide 1.97. Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident.
December 31, 1984.
10-7
.. w u;
.t.
o 10.2.1.6.14 IHP 1.100-1-2. Regulatory Guide 1.100. Seismic Qualification of Electrical Equipment for Nuclear Power Plants. December 31, 1984.
10.2.1.6.15 IMP 1.102-1-1 Regulatory Guide 1.102. Flood Protection for Nuclear Power Plants. September 1. 1982.
10.2.1.6.16 INP 1.106-1-1 Regulatory Guide 1.106. Thermal Overload Protection for Electric Motors on Motor-operated Valves, September 1. 1982.
10.2.1.6.17 IHP 1.115-1-1 Regulatory cuide 1,115. Protection Against Low-Trajectory Turbine Missiles. December 31, 1981.
10.2.1.6.18 IMP 1.117-1-1 Regulatory Guide 1.117. Tornado Design Glassification. September 1. 1982.
10.2.1.6.19 IMP 1.124-1-1 Regulatory Guide 1.124. Service Limits and Loading Combinations for Class 1E Linear-TYPE Component Supports. September 1. 1982.
10.2.1.7 PGE-1043 Accident Monitoring Instrumentation Review for the Trojan Nuclear Plant.
e 10.2.1.8 PGE-1048. Inservice Testing Program for Pumps and Valves.
Second Ten-Year Interval.
j 10.2.1.9 PCF-1049. Inservice Inspection Frogram. Second Ten-Year Interval.
10.2.1.10 PCE-8010. Nuclear Quality Assurance Program.
10.2.1.11 Trojan Safety-Related List.
10-8 n
e 10.2.2 TECHNICAL MANUAL REFERENCES 10.2.2.1 M12-54, Auxiliary Steam Generator Feedwater Pumps P-102A and P-1028 (Rev. 4).
Bingham-Willamette Company.
10.2.2.2 M12-55 Auxiliary Steam Generator Feedwater Pump Diesel Engine Driver K-107B (Rev. 1).
Waukesha Motor Company.
10.2.2.3 M12-61, Auxiliary Steam Generator Feedwater Pump Turbine Driver K-107A (Rev. 1).
Terry Steam Turbine Company.
10.2.2.4 N14440-10 Electric Auxiliary Feedwater Pump P-102 (Rev. 2).
Bingham-Willamette Company.
10.2.3 TROJAN PLANT OPERATING MANUAL (POM) REFERENCES 10.2.3.1 Of f-Normal Instruction (ONI)-55, Op.eration of Electric AFP Supplied by EDG.
10.2.3.2 Plant Operating Test (POT) 5-1, Pump and Valve Inservice Test.
i 10.2.3.3 POT-5-2, Valve Lineups and Inservice Testing.
10.2.3.4 POT-5-3, System Performance and Valve Inservice Test.
10.2.3.5 POT-25-2e, Switches S826 and S841 - K609 and K633 Trains A and B (SI,S).
10.2.3.6 Plant Operating Instruction OI-8-2. Auxiliary Feedwater.
e 10-9
- a b
10.3 PCE/ TROJAN PRINTS. DRAWINGS AND SPECIFICATIONS 10.3.1 SPECIFICATIONS 10.3.1.1 Trojan Nuclear Plant Technical Specifications.
10.3.1.2 PGE Electrical Design Guide No. 1.
10.3.1.3 E-2 Electrical Equipment Environmental Qualification.
10.3.1.4 TM-014 Auxiliary Feedwater Pump P-182 and Motor.
10.3.2 PIPING AND INSTRUMEMTATION DIACRAMS (P& ids) 10.3.2.1 M-208 (Rev. 28), Main Steam System.
10.3.2.2 M-213 Sheet 1 (Rev. 30), Condensate and Feedwater System.
10.3.2.3 M-213 Sheet 2 (Rev. 8), Condensate and Feedwater System.
10.3.2.4 M-213 Sheet 4 (Rev. 0), Condensate and Feedwater System.
10.3.2.5 M-214 Sheet 1 (Rev. 37), Auxiliary Steam 3ystem.
10.3.2.6 M-218 Sheet 1 (Rev. 33)
Service Water System.
10.3.2.7 M-226 (Rev. 18), Diesel Fuel Oil System.
10.3.2.8 M-228 (Rev. 33), Makeup Water Treatment System.
i 10.3.3 LOCIC DIACRAMS 10.3.3.1 E-1133, Auxiliary Feedwater Pump P-102A Channel'"A".
i l
l 10-10
.o ;. o..
, e :.'.
.t
f o
a 10.3.3.E E-1133A, Auxiliary Feedwater Pump P-102A Channel "A".
10.3.3.3 E-1134, Auxiliary Feedwater Pump P-102B Channel "B".
I 10.3.3.4 E-1134A, Auxiliary Feedwater Pump P-102B Channel "B".
10..'. 3. 5 E-1150, Motor Operated Valves.
10.3.3.5 E-1159, Stoam Turbine Driven Auxiliary Feedwater Pump Steam Inlet Valves.
l l
10.3.3.7 E-1183. Service Water System Motor Operated Valves.
1 10.3.3.8 E-1266, Heating and Ventilation Fans, Auxiliary Feedwater Pump Room.
10.3.3.9 E-1266A Heating and Ventilation Fans Auxiliary Feedwater Pump Room.
10.3.3.10 E-1285, Auxiliary Feedwater Pump Valves.
10.3.3.11 E-2300, Start-up Auxiliary Feedwater Pump P-182.
10.3.3.'2 E-2316, Start-up Auxiliary Feedwater Pump P-182 Dischargo l
l Valves.
10.3.4 ELECTRICAL PRIllTS 10.3.4.1 E-1 Plant Sin 51e Line Diagram (Rev. 14).
10.3.4.2 E-32 12.47-kV System (Rev. 8).
10.3.4.3 E-33 4.16-kV System Sheet 1 (Rev. 12).
10-11 u.
w
10.3.4.4 E-34 4.16-kV System Sheet 2 (Rev. 11).
10.3.4.5 E-35 480-V Load Centers Sheet 1, Engineered Safety Feature System (Rev. 6).
10.3.4.6 E-37 480-V Motor control Centers, Sheet 1, Engineered Safety Features System (Rev. 21).
10.3.4.7 E-38 480-V Control Centers, Sheet 2, Engineered Safety Feature System (Rev. 20).
10.3.4.8 E-44 125-V, 250-V de, and 208/120-V Instrument and Preferred ac System (Rev. 19).
10.3.4.9 E-44A 125-V, 250-V de, and 208/120 Instrument and Preferred ac System (Rev. 0).
10.3.4.10 E-45 i20-V Preferred ac Panels Y11 Y13. Y22, and Y24 Sheet No, Rev.
1 of 13 12 2 of 13 15 5 of 13 13 7 of 13 13 11 of 13 0
12 of 13 0
10.3.4.11 E-47 125-V de Distribution Panels D10. D20 D30, and D40.
Sheet No, Rev.
I 3 of 14 7
6 of 14 8
4 9 of 14 9
12 of 14 7
10.3.4.12 E-54 480-V Motor Control Centers Sheet 3. Eng'ineered Safety Features System (Rev. 24).
10-12 i
f 6
a w.
4 10.3.4.13 E-55 480-V Motor Control Centers, Sheet 4, Engineered Safety Features System (Rev. 18).
10.4 VENDOR REPORTS. SPECIFICATIONS. AND MISCELLANEOUS REFERENCES 10.4.1 Westinghouse Commercial Atomic Power WCAP-9600, Report on Small Break Accidents for Westinghouse NSSS Systems.
10.4.2 WCAP-7898 Long Term Transient Analysis for PWRs.
10.4.3 WCAP-7909, MARVEL - A Digital Computer Code for Transient Analysis for a Multiloop PWR System.
10.4.4 Franklin Research Center Technical Evaluation Report (TER) C5257-296 Auxiliary Feedwater System Automatic Initiation and Flow Indication (Trojan Nuclear Plant),
March 3, 1982.
10.4.5 Bechtel PC Topical Report BN-TOP-2. Design for Pipe Break Effects.
10.4.6 Bechtel Specification M-12, SCI Auxiliary Feedwater Pumps. Drivers, and Auxiliaries.
)
10.4.7 Bechtel Specification M-301, Specification for Pipin6 Materials and Standard Details, Rev. 1.
10.4.8 Bechtel Specification M-541. Sheets
- .1, 12, 13, control Valve Data Sheets, Rev. 4 I
l J
10.4.9 Bechtel Specification M-545-1 Pressure Safety Valve Data Sheet Rev. 2.
2 I
10.4.10 Bechtel Specification M-12-53, Auxiliary Feedwater Pump Diesel Engine Connection Diagram, Rev. 6.
I 10-13
' i. n e
' S EL
- - - ?
~
4 10.5 CORRESPONDENCE REFERENCES 10.$.1 VENDOR CORRESPONDENCE 10.5.1.1 Westinghouse Letter to Bechtel, POR-113. June 3, 1969 Auxiliary Feed Pumps.
10.5.1.2 Westinghouse Letter to Bechtel, POR-121, June 4, 1969, Auxiliary Feed Pumps.
10..
1.3 Westinghouse Letter to Bechtel, POR-151, July 22, 1969, Plant Startup and Shutdown.
10.5.1.4 Westinghouse Letter to Bechtel, POR-519. August 6, 1970 Auxiliary Feedwater System.
10.5.1.5 Westinghouse Letter to Bechtel, POR-728, March 29, 1971 Condtnsate Storage.
10.5.1.6 Westinghouse Letter to Bechtel, POR-751, April 14, 1971, Auxiliary Feedwater Initiation Time.
10.5.1.7 Wostinghouse Letter to Bechtel, POR-1472, October 20, 1972. Steam Systems Design Manual Transmittal.
10.5.1.8 Westinghouse Letter to Bechtel, POR-1613 February 14, 1973, Auxiliary Faedwater System Pump Capacities and Assured Storage Quality.
10.5.1.9 Westinghouse Letter to Bechtel, POR-1710, April 25, 1973 Auxiliary Feedwater System.
10.5.1.10 Westinghouse Letter to Bechtel, POR-1888 September,27, 1973 Auxiliary Feedwater System.
10-14 4-
e u
l i
10.5.1.11 Westinghouse Letter to Bechtel, PCR-2111, June 11, 1974 Auxiliary Feedwater Systam.
10.5.1.12 Westinghouse Letter to PCE, POR-85-579, May 24, 1985, Condensate Storage Tank Volume Requirements.
10.5.1.13 Bechtel Letter to PCE, BP-8650, November 1, 1977 RDC 77-139, Startup Auxiliary Feedwater Pump.
10.5.1.14 Bechtel Letter to PCE, BP-10485, July 11, 1980, RDC 80-003, Auxiliary Feedwater ac Independence.
10.5.1.15 Bechtel Letter to PCE, BP-10570, August 20, 1980 Motor-Driven AFWP Separate Flow Path.
1 10.5.1.16 Bechtel Letter to PCE, BP-12373. February 24, 1986, CST Level Requirements.
l 10.5.1.17 Bechtel Letter to PCE, BP-12422 March 27, 1986 ATW Suction Piping Analysis.
10.5.1.18 Bechtel Telephone Call to PCE, December 30, 1986, Verification of AFW Setpoints and Operation.
10.5.2 REGULATORY CORRESPot:0ENCE 10.5.2.1 Trojan Nuclear Plant Safety Evaluation Report (SEM),
i October 1974 Initial SER of Trojan Nuclear Plant.
10.5.2.2 KRC Letter to PCE, March 9, 1978, SER, Amendment No. 22 to KPF-1.
10.5.2.3 KRC Letter to PCE, October 3,1979, NRC Requirement's for Auxiliary Feedwater System at Trojan Nuclear Plant.
i 1
10-15
- 4
.c,.
e 10.5.2.4 PCE Letter to NRC, October 17, 1979, Response to NRC to PCE Letter Dated September 13, 1979.
10.5.2.5 PCE Letter to NRC, November 26, 1979, Response to NRC to PCE Letter Dated October 3, 1979.
10.5.2.6 NRC Letter to PCE, September 13, 1979 Follow-up Actions ResultinE from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident.
10.5.2.7 PCE Letter to NRC, December 31, 1979, Response to October 3,1979 NRC Questions Concerning the Trojan Auxiliary Feedwater System.
10.5.2.8 NRC Letter to PCE, March 25, 1980, SER, Auxiliary Feedwater System Requirements.
10.5.2.9 NRC Letter to PCE, May 14, 1980. Auxiliary Feedwater Sysham Requirements.
10.5.2.10 PCE Letter to NRC, July 25, 1980 Response to NRC to PCE Letter Dated May 15, 1980.
10.5.2.11 NRC Letter to PCE, October 23, 1980, SER. Implementation of Recommendt.tlons for Auxiliary Feedvater Systems.
10.5.2.12 POR Letter to NRC, December 12, 1980, Supplemental Response to SER Dated October 23, 1980, 10.5.2.13 NRC Letter to PCE, February 18, 1981 SER Supplement.
Implementation of Recommendations for the Auxiliary
/sedwater System.
e 10-16 1e
,-..v,
s 10.5.2.14 NRC Letter to PCE, August 4, 1981, SER, Seismic Qualification of the Airsiliary Feedwater System.
10.5.2.15 NRC Letter to PCE April 0, 1982, NUREC-0737 Item II.E.1.2, Auxiliary Feedwater System Automatic Initiation and Flow Indication.
10.5.2.16 NRC Letter to PCE, May 5, 1982, Testing of the Motor-Driven Auxiliary Feedwater Pump.
/
10.5.3 INTERNAL CORRESPOWDENCE 10.5.3.1 PCE Telephone Call to Westinghause, DIH-I22-80, September 5, 1980, Westinghouse Design Criteria for Auxiliary Feedwater Systems.
l 10.5.3.2 Memorandum ELD-03-82M, December 21, 1982, 125-V de Battery Capacity.
10.5.3.3 hemorandum RLS-1125-82M, December 16, 1982 Alternating Cur' tent Independence of Turtine-Driven Ard Pump Limitations due to Battery Capacity of Inverter Failues.
10.5.3.4 Memorandum 'dLS-03-83M, January 1. ?.983, hecommended Procedure Change to Maintain Cparable A Train 125-V de i
Battery.
10.5.3.5 Memorandum r.PY-44J-80, June 15,1983 Documentatiori of KPE-Recommer.ded Procedure Changes.
10.5.3.6 Memorandum ANR-172-84M, A. W. Roller to J. W. Lentsch, i
July 11, 1984 Verification Program Disposition Form j
Auxiliary Feedwater System Evaluation, NUREC-0737,.
Item II.E.1.1.
f 10-17 F-w,,
w
I l
b l
l 10.5.3.1 Memorandum RLS-1072-84M, R. L. Steele to J. W. Lentsch l
September 28, 1984, IE Information Notice No. 84-54.
i 10.5.3.8 Memorandum RLS-732-85M, R. L. Steele to J. W. Lentsch, June 5, 1985. Trojan Nuclear Plant Condensate Storage Tank (CST) Minimum Volune.
i l
10.5.3.9 Memorandum RLS-453-86M, April 16, 1986 Auxiliary l
Feedwater System Reliability Improvements.
i 10.5.3.10 Memorandum JWL-297-86M, May 28, 1986, Minimum Required CST Volume.
l 10.5.3.11 Memorandum RLS-691-86M, R. L. Steele to W. S. Orser, 1
June 24, 1986, CST Level Instrumentation and STS Minimum Volume.
l l
10.6 CALCULATION REFERENCES 10.6.1 PCE CALCULATIONS l
10.6.1.1 NSRD Calculation No. TNP-86-27 APW Flow Rate at End of l
CST Design Basis Volume, November 14, 1986.
10.6.1.2 MPEEB Crs1culation No, TE-028, 125-V de Station Battery capacity, November 10, 1962 (superseded).
10.6.1.3 NPEEB Calculation No.75-029, Statica Battery Required Load Shed During Statien B,lackout (superseded).
I 10.6.1.4 NPREB Calculation No. TE-105, 120-V ac Preferred Instrument Bus.
10.6.1.5 N,PEEB Calculation No. TE-115, AFW Pump / CST Level Control Setpoints, July 24, 1986.
10-18
'~~
O r$%.
} '*.,
o 4
10.6.1.6 NPEEB Calculation No. TE-119. Station Battery capacity, September 11, 1986.
10.6.1.7 NPEMB Calculation No. TM-002. AW Pump Overspeed Trip Test Discharge Pressure, January 10, 1981.
10.6.1.8 NPEMB Calculation No. TM-006, AFW System Steam Pressure.
[
January 10, 1981.
10.6.1.9 NPEMB Calculation No. TM-082, AFW Panel Room Ventilation, February 14, 1983.
10.6.1.10 NPEMB Calculation No. TM-158, AFW Reliability / CST Suction Piping, June 19, 1986.
I 10.6.1.11 NPEMB Calculation No. TM-185, Stress Analysis of AFW Turbine Exhetst Drain Line, November 13, 1986.
10.6.1.22 NPEN3 Calculation No. TM-191, Seismic Analysis of AFW Pump Lube Oil Drain PiF118 December 16, 1986.
10.6.1.13 MPEc8 Calculation No. TC-055.. AFW Pump Foundation Design.
November 5, 1985.
1 10.6.'. 14 NVECM Calculatten No. TC-258, FIS 3004 Al-D2 Rigid Lounting Configuration Rev. 1. June 5, 1985.
10.6.1.15 NPECB Calculation No. TC-265, Seismic Qualification of CV-3004 Replacement Valve Operators February 9,1985.
10.6.1.16 NPEca Calculation No. TC-296, Pipe Support Modifications.
September 11, 1985.
e 10-19
.s
o A
2 10.6.1.17 WPECB Calculation No. TC-335. Condensate Storage Tank Evaluation. April 17, 1986.
l 10.6.1.18 NPECB Calculation No. TC-336, CST Level Transmitter Enclosures, April 25, 1986.
L 10.6.1.19 NPECB Calculation No. TC-348, AFW Pump Reliability Improvements, June 2, 1986.
l 10.6.1.20 NPECS Calculation No. TC-378, AFW Room Exhaust Flappers, 4
October 23, 1986.
l 10.6.1.21 WPECB Calculation No. TC-380 C-160 Room Exhaust Ductwork, October 28, 1986.
10.6.1.22 NPECB Calculation No. TC-399, AFW Ductwork, December 15 j
1986.
10.6.1.23 Accumulator S1sint Calculation (RDC 80-003), March 31, 1981.
s 104.l lgQgft, CALCUt.ATt033 3
i 10.6.2.1 12-22, Fuel O!.1 Day tank T-152 Citian, September 26, 1975.
]
j 10.6.2.2 16-14, AFW Pump Turbine Drain System Flow Restrictions, December 28, 1976.
i l
]
10.6.2.3 16-17 AFW Pump Turbine Drain Condenser Sizing, April 29, 1977.
l I
10.6.2.4 16-42, APW Pump WPSH Available, June 23, 1978.
i 10-20
~~
~
A 10.6.2.5 16-43. P-182 Low Suction Pressure Trip 9etpoint, October 9, 1978.
10.6.2.6 16-44, Minimum CST Level for AFW Puep NPSH Requirements, July 3, 1979.
10.6.2.7 16-47, NPSH Available for Two AW Pump Operation.
April 22, 1980.
10.6.2.8 16-48, AFW Isolation Following Main Steam Line Break, June 30, 1980.
10.6.2.9 16-49, Maximum AFW Flow to Faulted Steam Generator, 1
March 26, 1982.
I 10.6.2.10 16-50, Safety-Related MOV Design Basis, November 1,1986, i
i 10.6.2.11 16-52, AFW Systan Pressure due to Pump Overspeed.
l t
10.6.2.12 AFW Punps, Check AFWP Sizing in Relation to As-Built l
1 System. September 17, 1969.
l l
l 4
t l
i t
I i
10-21 2
a '.
i
C
+
a APPgNDIX A OPgN ITgMS The following items require some level of resolution to be incorporated as part of the design basis for the Auxiliary Feedwater system.
1.
The assumed maximum steady state flow rate for internal flooding for the AFW system needs verification.
Design flow rate for each Circulating Water System train is 210.000 spm; but some documents Laply an assumed flooding rate of 500.000 spm.
(Section 3.7.1) 2.
The design basis for piping pressure and temperature ratings throughout the system needs further explanation.
Specification M-301 is referenced for actual pressure and temperature ratings but the functional basis for selection of required parameters for each piping spool is not available.
Analysis of piping classification code schedules, system P& ids, and some walkdowns are necessary to resolve.
(section 3.1) 3.
The following Bechtel calculations are needed to support design basis informationt 12-22, ruel oil Day Tank Sising. 9/26/75.
a 16-14. AFW Pump Turbine Drain System Flow Restrictions. 12/28/76 16-17. AFW Pump Turbine Crain Condenser Sizing. 4/29/77 16-49 Maximum AFW Flow to Faulted SG 3/26/82.
4.
The design margin included in AFW System flow rate requirements is not '
available. Accident analyses assume a sinimum available flow of 426 spm, while Westinghouse-supplied design criteria (and, specifically.
Westinghouse Intters to Bechtel POR-113. dated June 3. 1969, and P03-121, i
dated June 4. 1969) require a minimum flow of 440 spm.
Any basis for the margin bstween these two figures, oc for rny margin includea in the e
seeident knalysis flow rate of 426 gra, should be provided by l
(Secticn 4.1.3)
[
5.
Calculations uopporting the basis for the change in September 1986 to the X-107A trip set points need verification.
Trojan plant Engineering po'.'for9ed calculations to support the change, but these have not been i
incorporated into Npg or NSRD controlled calculation filas.
(Seccion
[
4.2.2)
[
\\
6.
Diesel driver X-1078 design horsepower is known, but an evaluation As l
needed to determine actual pump horsepower required for proper operation i
of P-1028 at rated speed.
(Section 4.3.2) 7.
A eticulation is needed (Bechtel Calculation 12-22. Puel Oil Day Tank f
Sizing. 9/26/75, may suffice when received) to verify the 91 sing criteria for diesel fuel oil day tank T-152.
This is necessary to confirm that the tank capacity of 500 gallons and operability requirement of 450 gallons s
are based on K-1078 fuel consumption requirements. (Sections 4.3.2. 4.3.3) i A-1 t
k A
.o
.i
o 8.
High AFW flow in a faulted branch line will result in isolation of that branch at 500 spa. NQAD has expressed a concern that the basis for this flow rate selection be verified.
Considering the AFW pump rating of 880 spm, less 500 spa lost through the faulted branch line, would leave only 380 spa deliverable to the steam generators.
This flow rate would be less than the 426 spa assumed in the worst-case accident analyses.
Bechtel Calculation 15-48, AFW Isolation Following Main Steam Line Break, Juns 30, 1980, does not resolve this item, and further evaluation is needed.
Bechtel Calculation 15-49, Maximus AFW Flow to Faulted SG, March 26, 1982, may be germane when received.
(Section 4.6.2) 9.
Westinghouse letter POR-1613 to Bechtel dated February 14,1973, and Curve SSg-1119 established the minimum usable CST volume requirement of 196,000 gallons.
However, no Westinghouse calculations are available regarding the basis for determining this value.
Required CST volume assuming rated AFW pump flow for two hours at hot standby followed by four hours of cooldown far exceeds 196.000 gallons.
It would appear that the average flow required over the six hour period does not exceed 544 spm.
An evaluation is needed to determine assumptions for flow requirements at various stages during plant cooldown.
NSRD calculation TNP-86-27, AFW Flow Rate at Jnd of CST Design Basis Volume, dated November 14, 1986, refers to this item but does not resolve it.
(Section 5.0)
- 10. Diesel driver K-1078 battery sizing design basis needs to be verified.
RDC 80-086 is pertinent.
(Section 4.3.2)
- 11. Westinghouse letter POR-751 to Bechtel dated April 14, 1971, established a requirement that the AFW pumps be operating at rated speed and deliv2 ring rated flow within one minute following actuation of any automatic signal which starts the pumps. The design basis assumptions or calcu14tions supporting the figure of one minute are needed.
Alsts needed t.re any margin calculations or assumptions ircluded, (Section 1.2)
- 12. Design basis for the 100 psid ret point for AFW pump dirferential pressure control is unavailable.
Determination would require Westing 5ouse input.
(Section 4.11.2)
- 13. Design margins for cooling water to AFW pumps F-102A and P-1028 auxiliaries are unavailable.
Resolution will require pump vendor input.
(Section 4.1.3) l t
- 14. Basis for diesel driver K-107D lube oil cooler cooling water requirements are unavailable.
Resolution will require Waukasha input.
(Section 4.3.2)
{
l
- 15. Design basis, if any, for AFW pump P-182 flow rate and re2irculation flow rate characteristics are unavailable.
(Section 4.4.2)
- 16. A calculation is needed to support the design basis for sizing of i
accumulators T-166A through T-166D.
(Section 4.7.1.1)
- 17. Design basis for valve M0-3170 stroke time is needed.
(Section 4.7.1.2) l l
l A-2 i
i
- .,,4e.
+~<m 1
.