ML18026A308

Forwards ASME Boiler & Pressure Vessel Code,Section 3 Overpressure Rept Per FSAR Question 211.204

ML18026A308
Person / Time
Site:	Susquehanna
Issue date:	09/12/1980
From:	Curtis N PENNSYLVANIA POWER & LIGHT CO.
To:	Youngblood B Office of Nuclear Reactor Regulation
References
ER-100450, PLA-544, NUDOCS 8009170290
Download: ML18026A308 (19)

Text

ACCESS IOIJ MT3R F ACIL:50

'387 50-388 AUTH ~ IJA+E CURTISsN ~ A' R E C I P ~ V. 4! 1E YQLNG8LQOO ~ g ~ J ~

REG )LA 'lRY I')FOR<AT 10

(R IQS)

8009170290 DQC ~ OA t'E: 80/09/12

'lO I >R I ZEO!

~0 OOCKET Susauehanna Steam Electric Stations Jnit ip Pennsylva 05000387 Susouehanna Steam Electric Station, Jnit 2E Pennsylva 05000388 AUTHOR AFF ILI 4 T I 0 Pennsylvania Po<<er K Lioht Co ~

REC IP IEMT AFF ILI 4 T IO,'l Licensina 3ranch 1

SUR JECT:

Fo ( wa rds AS'3E 8o i l er 8

P ressure Yesse 1

Co:jes Sec t i on '3 overores sure r eot oer FSAR Quest i on 211

~ 20<< ~

33ISISISITTIO~ CSE:

33001S COPIES SECEIVED:LTS ji'3CL /'IZ~:

T I TLE:

PSAR/FSAR 4 4DTS and Re l ated Corr esoondence

'lOTE,S:Send ICE 3 cooies FSAH 'll amends

~

Send Is33E 3 cooies FSAR 4 all aTIsenCis.

05000387 05000388 ACT IO~:

RFC I>IE'lI 10 COOEIVANE 4/0 LICE>>8VG.

PUSHBROO<E'4

~

COP IES L T TR Es'iCL 0

1 0

REC I P I E'II IO. COOE/'i 4'<E YOIJ'}33BLOOG,:l STARK,P, pg COPIFS L T TR E'IICL 1

0 1

1 I "iTER'44L'.

ACCI 0 EYAL RR?6 E'IG 9R 0

CORE

>ERF Tlk 10 EvFRG GE )SC IE'~CES 1'l HYD/G 3

3H 15 ICE 06 L I C 0 JAL B.l

"<EC'3 " 'lG BR 1 8 OP LI C 9R PROC/TST REV 20

""SS IR22 "FG F ILF 01 S T RUC T E '46 8R25 1

1 1

1 1

1 1

0 1

1 1

1 3

3 1

1 1,

1 1

1 1

1 1

1 1

1 1

1 1

1 AUX SYS 9R 9 7 COiT SVS qR ng EFP TP 8YS LlR12 EAU I P

'PUAL 89 1 3 HV'" F ACT E 'IG 9R ICC SYS 9P j103 LI C GUIO

<<TL E'uG PR 1 7 vPA OELO POWER SYS (lA "HR 21 REAC SYS BR 23 SI l 4'4AL 8R 24 SYS I "JTEPAC 9R 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 1

1 0

0 1

1 1

1 1

E X TER".'AL:

ACRS t> 8 tgsO T'lTAL 'LLT""t<

J~

L s

J 4

wc33 J I t 4

~

LT TR sl C f

TWO NORTH NINTH STREET, ALLENTOWN, PA.

18101 PHONE:

(215) 821 ~ 5151 NORMAN W. CURTIS Vice presroenr Engineering 8 Consrrucrion 82 I-5381 September 12, 1980 Nr. B. 3. Youngblood, Chief Licensing Branch No.

1 Division of Licensing U.ST Nuclear Regulatory Commission Washington, DC 20555 Docket Nos..

50-387 50-388 SUSQUEHANNA STEAN ELECTRIC STATION FSAR QUESTION 211.204 ER 100450 PILE 841-2 PLA-544

Dear Nr. Youngblood:

As requested in PSAR Question 211.204, attached is a copy of the AS~r1E Boiler and Pressure Vessel Code,Section III overpressure report for Susquehanna.

Very truly yours,

~~

~ 7/ZyLg~

g N.

W. Curtis Vice President-Engineering and Construction-Nuclear CTC:mks Attachment i'0091 0 Q~~

PENNSYLVANIA POWER 8

LIGHT COMPANY

NUCLEAR ENERGY DIVISION Abstract This report provides sufficient information and documentation to show compliance with all requirements of Article 9 of ASHE Pressure Vessel Code - Section I'II, 1968, Nuclear Vessels (up to and including Summer 1970 Addenda) in the area of the vessel overpressure protection design of the Susquehanna 1

and Susquehanna 2 nuclear pressure vessels.

The effects on the vessel pressure transients of'alve capacity are also shown.

NUCl.EAR ENERGY DlVlslON Table of Contents 1.

Introduction 2.

Design Basis 3.

Method of Analysis 4.

System Design 5.

Evaluation of Results 6.

Safety/Relief Valve Characteristi cs 7.

Concl us ions

GENERALO ELECTRIC NUCLEAR ENERGY DIVISION List of Illustrations Figure 1

Figure 2

Figure 3

Figure 4

Figure 5

Figure 6

Figure 7

Typical S/R Valve Capacity Characteristic SCRAM Reactivity vs Time SCRAM Rod Drive vs Time Peak Vessel Pressure vs Safety/Relief Valve Capacity Time Response of Pressurization Transients Sa fety/Relief Valve Schematic Elevation Safety/Relief Valve Schematic Plan

GENERAL ELECTRlC NUCLEAR ENERGY DIVNEOM 1.

INTRODUCTION 1.1 The vessel overpressure protection system is designed to satisfy the require-ments of Section III, Nuclear Vessels, of the ASHE Boiler and Pressure Vessel Code.

The general requirements 'for protection against overpressure as given in Article 9 of Section III of the Code recognize that reactor vessel overpressure protection is one function of the reactor protective systems and allows the integration of pressure relief devices with the protective systems of the nuclear reactor.

Hence, credit is taken for the SCRAH protective system as a complementary pressure protection device.

2.

%SIGN BASIS 2.1 Safety Valve Caoacit The safety valve capacity of this plant is adequate to limit the primary system pressure, including transients, to the requiremnts of the ASHE Boiler and Pressure Vessel Code,Section III, 1968, Nuclear Vessels (up to and including Summer 1970 Addenda).

The essential ASHE requirements which are all met by this analysis are:

2.1.1 It is recognized that the protection of vessels in a nuclear power plant is dependent upon many protective systems to relieve or terminate pressure transients.

Installation of pressure relieving devices may not independently provide complete protection.

2.1.2 The safety valve sizing evaluation assumes credit for operation of the SCRAH protective system which may be tripped by any one of two sources; i.e.,

a direct, or flux signal.

The di rect SCRAM'ignal is derived from position switches mounted on the main steamline isolation valves or the turbine stop valves or from pressure switches mounted on the dump valve of the turbine control valve hydraulic actuation system.

The position switches are actuated when the respective valves are closing and following 10 percent travel of full stroke.

The pressure switches are actuated when a fast closure of the control'alves is initiated.

Further, no credit is taken for power operation of the pressure relieving devices.

Credit is taken for the dual purpose safety/relief valves in their ASPIC Code qualified mode of safety operation.

S~II ~a~I. i

<LCCTaIC NUCI.EAR ENERGY DIVISION 2.1.3 The nominal'ressure setting of at least one safety/relief valve connected to any: vesse 1

or system shall not be greater than a pressure at the safety/relief valves corresponding to the design pressure

{1250 psig) anywhere in the rotected p

vessel.

2.1.4 The rated capacity of the pressure relieving devices shall be sufficient to prevent a rise in pressure within the protected vessel of more than 110 percent of the design pressure

{1.10X1250psig 1375 psig) for events defined in Paragraph 4.3.1.

2.1.5 Full account is taken of the pressure drop on both the inlet and discharge sides of the valves.

All combination safety/relief va1ves discharge into the suppression pool through a discharge pipe from each valve which is designed to achieve sonic flow conditions through the valve; thus providing flow independence to discharge piping losses.

3.

KTHOD OF ANALYSIS 3.1 To design the pressure protection for'the nuclear boiler system, extensive analytical models representing all essential dynamic characteristics of the system are simulated on a large computing facility.

These models include the hydrodynamics of the flow loop, the reactor kinetics, the thermal characteristics of the fuel and its transfer of heat to the coolant, and all the principa controller features, such as feedwater flow, reciI culation flow, reactor water level, pressure, and load demand.

These are represented with all their principal

'on-linear features in models that have evolved through extensive experience and favorable comparison of analysis with actual BWR test data.

3.1.1 A detailed description of this model is documented in licensing topical report NED0-10S02, Analytical Hethods of Plant Transient Evaluations for the K-BH,, R.B. Linford.

Included within this model, then, are components of the reactor vessel pressure protection system, which system is the subject of this repor Dual safety/relief valves are simulated in the non-linear representation, d th model thereby allows full investigation of the various valve response

times, valve capacities, and actuation setpoints that are available in applicable h

~

~

ard-mre systems.

3.1.2 Typical capacity characteristics's modeled are represented in Figure 1

for the safety/relief valves.

The 'associated

bypass, turbine control valve, and mainsteam isolation valve charac eristics are, of course, also represented-fully in the model.

GENERAL I ELECTRIC NUCLEAR ENERGY DIVISION 4.

SYSTEM DESIGN 4.1

.A parametric study was conducted to determine the required steam flow capacity of the safety/relief valves based on the following assumptions.

4.2 Ooeratin Conditions 4.2.1 Operating Power 3439Nt (104.4 percent of reactor rated power) 1020 psig 5 teamf1 ow 14.153xlOs lbs/hr (105 percent of rated steam flow)

These conditions are the most severe because the maximum stored energy exists at these conditions.

At lower power conditions the transients would be less

severe, Vessel Dona Pressure 4.4 Scram a.

SCRAM reactivity curve Figure 2

b.

Control rod drive SCRAM motion Figure 3

4.5 Safety Relief Valve Transient Analysis Specifications 4.3 Transi ents 4.3.1 The overpressure protection system must accomnodate the est severe pres-suriza.ion transient.

There are two major transients, the closure of all main steam line isolation valves and a turbine/generator trip with a coincident failure of the turbine steam bypass system valves to open, that represent the most severe abnormal operational transients resulting in a nuclear system pressure rise.

The evaluation of transient behavior with final plant configuration has shown that the isolation valve closure is slightly more severe when credit is taken only for indirect'erived scrams, therefore, it is used as the overpressure protection basis event.

a.

Valve groups 5

b.

Pressure setpoi nt 3165 - 1205 psig

(+1 percent assumed error) 4.6 Sa etv Valve Caoaci 4.6.1 Sizing of the safety valve capacity is based on establishing an adequate margin from the peak vessel pressure to the vessel code limit (1375 psig) in response to the reference transients.

NUCLEAR ENERGY DIVISION 5.

EVALUATION OF RESULTS 5.1, Safet Valve Ca acit 5.1.1 The parametric relationship between peak vessel (bottom) pressure and safety valve capacity for the HSIV transient with high flux and position trip scram is described in Figure 4, Also shown in Figure 4 is the parametric relationship between peak vessel (bottom) pressure and safety valve capacity for the turbine trip with -a coincident failure of the turbine bypass valves to open and direct scram, which is the most severe transient when direct scram is considered.

Pressures shown for flux scram will result only with multiple failure in the redundant direct scram system.

5.1.2 The time response of the vessel pressure to the MSIV transient with flux scram and the turbine trip with a coincident failure of the turbine bypass valves to open and direct scram for 16 valves is illustrated in Figure 5, This shows that the pressure at the vessel bottom exceeds 1250 psig for less than 6 seconds which is

'ot long enough to transfer any appreciable amount of heat into the vessel metal which was at a temperature well below 550'F at the start of the transient.

5.1.3 From the analytical models described in Paragraph 3 together with engineering studies, it has been determined that the safety/relief valve reclosing pressures, as specified in Paragraph 6.3.1, are acceptable.

6.

SAFETY/RELIEF VALVE CHARACTERISTICS 6.1 Schematic Arrangement.

The schematic arrangement of the safety/relief valves are s

own 1n figures 6 and 7.

6. 2 Pressure Drop in Inlet and Di s charge 6.2.1 Pressure drop on the piping from the reactor vessel to the valves is taken into account in calculating the maximum vessel pressures reported above.

6.2.2 Pressure drop in the discharge piping to the suppression pool is limited by proper discharge line sizing to prevent back pressure on each safety/relief valve from exceeding 40 percent of the valve inlet pressure, thus assuring choked flow in the valve orifice and no reduction'of valve capacity due to the discharge piping.

Each safety/relief valve has its own separate discharge line.

GENERALi EI.ECTRIC MUCLEAR ENERGY DIVISION 6.3 Safet /Relief Valve Oescri tion 6.3.1 These valves were manufactured by Crosby Valve and Gage Company to ASHE, Secfion III Code, 1971 Edition.

They comply with ASHE III, Paragraph NB-7640 as safety valves with auxiliary actuating devices.

guantities and set points are as follows; uantit 7.

CONCLUSION Opening Set Point

~SI 1146 1175 1185 1195 1205 Reclosing Pressure

~SI 1020 1046 1055 1064 1072 ASHE Rated Capacity at 103 percent of Set Pressure lb/hr minimum 862,400 8830950 891,380 898,800 9060250 7.1 Safety requirements have long demanded very high reliability in the reactor SCRAH functions.

Recognition of this reliability as being completely adequate justification for these functions to contribute to vessel pressure protection is reflected in the Section III Code provisions.

Actual General Electric design practice very conservatively applies the code provisions which results in margins even beyond those necessary to satisfy code limits which further enhances the reliability of vessel pressure protection;

SEN ERALe Kl!cTlllC NUCLEAR RNKRQY CKVNlON

~

~

~

~

I

's I

~

~

~

I

~

s t

I I

I I

~

t 100

~

I

~

I

~

~

~

I I

~

~

SO

~

~

s s

I

~

~

~

~

I

~

s I

~

I I

s ASME CODE APPROVED

~ 'AFETY VALVE CAPAC ITY

~

~

~

~

~

~

~

~

~

~

~

~

~

s t tel ill I

I '

I

~

I

~

60

~

~

I I

I I

~

~

I

~

~

ASSUMED LIFT C }{ARACTER1 ST IC

~

s

~

~

~

I s

~

~

I

~

s I

40 I

~

20

~

I

~

~

I I

~

I

~

~

s, I

I I

~

I

~

I I

I '

~

I

~

~

\\

I

~

~

~

I

~

I I

~

~

I f

~

~

~

~

~

~

~

~

I I

~

~

~

I

~

~

~

~

I I

~

t

~

I

~

~

~

I I

~

I I

~

I

~

~

~

~

s

~

I s

~

~

~

~

I

~

~ '

s

~

t I

~

~

I

~

I I

~

I

~

~

I

~

I

~

I I

0

~

I I

~

~

s

~

.96

.97

. 98

.99 1.00

).01 1.02 1.03 1.04 INLET PRESSURE /

SET POINT PRESSURE.

FIGURE l TYPICAL S/R VALVE CAPACITY CHARACTERISTIC

REv. eo.

1 GENERA

'LECTRIC NUCLEAR ENERGY DIVISION 40 30 NOMINAL EOC-l SCRAM CURVE SCRAM CURVE USED IN ANALYSIS lo 2

3 4"

TIK AFTER SCRAM (SECONQS)

Figure 2

SCRAM Reactivity Versus Time 5

GENERALELERIC NUCLEAR KNfRGV DEVISlON 110 100 80 60 Ch CL I

CO gp 20 I

3 TIN AMiR SCRAM (SECONDS)

Figur 3

SCRAM Rod Orive Versus Time

GEH!RALELEOhLC NUCLEAR ENERGY DlV1NOhl 1450 1375 PSI CODE LIMIT 1350 Ih ALL MSIV CLOSURE, FLUX SCRAM 1300 TURBINE TRIP, W/0 BYPASS TRIP SCRAH 1250 ALL HSI V CLOSURE, POSITION SCRAM I

1 2

l 3

4 5

I I

I 6

7 8

9 10

)1 12 13

) 4 15 16 17 18 19 NUKER OF SAFETY VALVES Figure 4 Peak Vesse1 Pressure Versus Safety/Relief VaIve Capacity

GENEHALELEcT!L NlJCLXAR ENERGY DlVIQCW 1400 1300 ALL MSIY CLOSURE, FLUX SCRAM 1200 TURBINE TRIP M/0 BYPASS TRIP SCRAM 1100 I

8 I

2 4

6 TIW (SECO~DS)

Figure 5

Time Response of Pressurization Transients 10

GENERAL! LEGTlO NUCLEAR KNKRQYDIVISON ORYWELL REACTOR VESSELS.

HAIN STEAM LINE N!N STEAM

!SQLATION VALV S SAFr.TY/

RELIEF VALVES FLOW RESTRICTOR SUPPRESSION POOL Figure 6

Safety/Relief Valve Schematic Elevation

GEMERALELKCTRlc NUCLEAR ENERGY blVtOON NIN STEAN LINES NIN STEAN ISOLATION VALYE5 DRYWELL S/R Y

S/R V

S/R V

S/R Y

S/R V

S/R V

REACTOR. VESSEL S/R V

S/R V

S/R V

S/R V

S/R V

S/R V

S/R V

S/R V

S/R Y

S/R Y

Figure 7

Safety/Relief Valve Schematic Plan