ML20079E779

Rev 0 to Overpressure Protection Rept

ML20079E779
Person / Time
Site:	Hope Creek
Issue date:	01/28/1983
From:	Ikemoto F, Mata M, Nazareno E GENERAL ELECTRIC CO.
To:
Shared Package
ML20079E750	List: ML20079E748 ML20079E779 ML20079E792
References
23A1411, 23A1411-R, 23A1411-R00, NUDOCS 8401170420
Download: ML20079E779 (14)

Text

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B STRACT This report provides sufficient information and doc,amentation to show compliance with all requirements of Article 9 of ASME Pressura Yessel Code - Section III. Nuolear Vessels 1968 Edition with Addenda to and including vinter,1969 Addenda in the aras of the vessel overpressure protection design of the Hope Creek Unit i nuclear pressure vessels.

l The effects on the vessel pressure transients of valve capacity are also

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TABLE OF CONIENTS

1. INTR (BUCTION i 2. DESIGN BASIS i

3. METE (B 0F ANALYSIS I

i 4. SYSTEM DESIGN

5. EVALUATION OF RESW,75 l 6. SAFRTI/ RELIEF YALVE CHARACTTRISTICS I'

7. CONQ,USIONS

, 8. REFERENCES List of Finures t

1 Typical S/R Valve Capacity Characteristio 2 SCRAM Rod Drive vs Time i

. 3 Peak Yessel Pressure vs Safety / Relief Yalve Capacity 4 Tine Response of Pressurization Transients J e

5 Safety /Relle'f Valve Schematic Elevation .

6 Safety / Relief Yalve Schematio Plan k

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1. INIRODUCTION 1.1 he vessel overpressare protectica system li, designed to satisfy the requirements of Section III, Neoloar Yessels, of the ASIE Boiler and Pressure Yessel Code. The general requirements for protection against overpressure as gives in Article 9 of Section III of the Code recognise that resotor vessel overpressure protection is one fanation of the reactor protective systems and allows the integration of pressare relief devices with the protectivs systems of the nuclear reactor. Hence, credit is taken for the sorsa prctactive system as a complementary pressure protection device.

2. DESIGN B ASIS 2.1 Safetv/ Relief Valve Cansoitv. The safety / relief valve capacitys of this plant is adeguate to limit the primary system pressure, including trauients, to the requirements of the ASME Boiler and Pressure Yessel Code,Section III, Neelear Yessels 1968 Edition with Addenda to and including Winter,1969 11denda. The essential ASME 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 Mt Independently provide complete protection.

]n 2.1.2 The safety / relief valve sizing evaluation assumes credit for operation of the scram protective system which may be tripped by either of two sources; le, a direct, or finz signal. The diroot scraa signal is derived from position switches mounted on the main atesaline 1. solation valves, or the turbine stop valves, or from pressure switches mounted on the dump valve of the turbine control valve hydraulio 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 fas.t closure of the oostrol valves is initiated. Credit is taken for the dual purpose safsty/ relief valves in their ASME Code qualified mode of safety operatior.,

2.1.3 The nominal pressure setting of at least one safety / relief valve connected to any vessel or system shall not be greater than a pressure at the ,

safety / relief valves corresponding to the design pressure (1250 psig) anywhere in the protected vessel.

2.1.4 The rated capacity of the pressure relhving devices shall be sufficient to prevent a rise in pressure vitain the protected vessel of more than 110 percent of the design pressate (1.10 .~ 1250 psig = 1375 psig) for events defined in Paragraph 4.3.1.

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2.1.5 Full aseoast is takes of the pressare drop on'both the inlet and discharge sides of the valves. All sembination safety / relief valves discharge into the suppression pool through a discharge pipe from each valve which is designed to achieve somie flow conditions through the valvei thus providing flow independence to discharge piping losses.

3. MB1EOD OF ANALYSIS 3 .1 To design the pressure protection for the maaleur boiler system, <

extensive analytzcal models representing all essential dynamic characteristics of the system are simulated on a large computing facility. These models Asclude the hydrodysesios of the flow loop, the reactor kinetics, the thermal characteristics of the fuel and its transfer of heat to the coolant, gad all the principal controller features, such as feedwate.r flow, rectroslation flow, remotor water level, pressure, and load demand. Th6se are represented with all their principal moalinear festares in models that have evolved through extensive experience and favorable comparison of analysis with notaal BTR test data.

3.1.1 A detallad description of this model (Paragraph 8.2.a) is documented as a licensing topical report, Included within this model, them, are son-ponents of the reactor vessel pressure protection system, which is the subject of this report. Dual safety / relief valves are simulated in the monlinear (p.'.

representation, sad the model thereby allows full investigation of the various valve response times, valve espacities, and motsation setpoints that are available in applicable hardware systems.

3.1.2 Typical espacity characteristics as modeled are represented in Figure 1 for the safety / relief valves. The assoolated bypa'ss, turbine control valve, and asia steam- isolation valve characteristics are, of course, also represented fully in the model.

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4. SY3'IEN DESIGN 4.1 A parametrie study was sondseted to determine the required steam flow espacity of the safety / relief valves based on the following assumptions. i 4.1 Onoratian conditions 4.2.1 Operating Power -

3435 MWt (104.3 percent of uselear

, rated power)

Vessel Domme Pressure -

,{,1020 psis Steamflor -

14.864 x 106 lb/hr (105 percent of nuclear boiler rated steam flow)

These cor.11tions are the most severe because the naziram stored energy esists at these conditions. At lower power conditions the temasients wonid be less severe.

4.3 Transi tala.

l 4.3.1 The overpressure protectica systen must accommodate the most severe '

l pressurization transient. There are two major transients: the closure of all main steam line isolation valves, and a turbine / generator trip with a 3'

ooinci' '; failure of the turbias steam bypass system valves to open, that r ep-- .: the most severs abnormal operational transients resulting in a nucleat- system pressure rise. The evaluation of transient behavior with final plant configuration has shown that the isolation valve closure is sligt.tly more severe when credit is taken only for indirect derived scras.; therefore, it is used as the overpressure protection basis event.

4.4 .hzam .

a. Scram reactivity carve

• l l b. Control rod drive Scram motion - Figure 2 l

4.5 Safetv/Retief Yalve Trangient Analysis Sneelfications

a. Yalve groups -

3

b. Pressure setpoint -

1121 to 1141 psig (includes - + one percent assnaed error)

• Scram rasetivity of each transiest analyzed is calculated within the ODYN Computer Program ..-

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5. EVALUATION OF RESULTS 5.1 Safetv/ Relief Yalve canseity 5.1.1 ne parametric relationship between peak vessel (bottom) pressure and

' safety / relief valve capacity for the MSIV closure transient with high sentron

. fluz scram is described in Figure 3. Also shows in Ngure 3 is the parametric -

l relationship between peak vessel (botton) pressure and safety / relief valve I 1 capacity for generator losa rejection with a coincident failure of the turbine J

bypass valves to open and direct scram, which is the most severe transient l

when direct scram is considered. Pressures shown for fluz scram will result

, only with multiple failure in the rednadant direct scram system.

5.1.2 ne time response of the vessel pressure to the MSIY transient with fluz screa and the generator load rejection with a coincident failste of the turbine bypass valves to open and direct scram for 14 valves is illustrated in

e. Figure 4. His shows that the pressure at the vessel bottom exceeds 1200 psig

(' for less than 5.0 seconds, which is not long enough to transfer any appreciable amount of heat into the vessel metal, which was at a temperature well holow 550*F st the start of the transient.

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

6. SAFEIT/ RELIEF YALVE GARACIERISTICS ,

6.1 Sehenstic Arrannement. He schematic arrangement of the safety / relief valves are shown in Figures 5 an.16. -

6.2 Pressure Dron ,ig Jgist and Discharme ,

6.2.1 Pressure deep on the riping from the reactor. vessel to the valves is taken lato account in calenlating 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

., safetv/ relief valve from exceeding 40 percent of the valve inlet pressure, thus assasing choked flow in the valve orifice and no redaction of ulve capacity due to the discharge piping. Each safety / relief valve has its own

separate discharge line.

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6.3 Jgigtv/ Relief Yalve Descrintion 6.3.1 nose valves were mamafectured by Target Roc,k Corportion to ASME, '

Section III Code,1968 Edition (up to and including Summer 1970 Addenda).

They comply with ASME III, Paragraph N911.4 for pilot-operated valves. His report justifies the use of setpoints up to the following.

Opening Minimum Reclosing ASME Rated Capacity at Setpoint Pressure 103 Percent of Set Pressure Susatity _psin nain Ib/hr minimum

~. 4 1108 9 86 884,000 5 3120 997 893.000 5 1130 1007 901,000 5

7. CONG.USION 7.1 Safety requirassats hare long demanded very high reliability la the reactor Suas functions. Recognition of this reliability as being contpletely adequate justifloation for these fanations to contribute to vessel pressure protection is reflected la the Section III Code provisions. Actual General Risotrio design practica very conservatively applies the code provisions which results la margins even beyond those necessary to satisfy code limits which further enhances the relisbility of vessel pressure protection. ']-

8. REFERENCES 8.1 American !:ociety of Meeheniesi Enmineers. ASME Pressure Yessel Code, '

Section III - Nuclear Yesse.) s.

8.2 General Electric Documents

a. Qualification of One-Dimensional Core Transients of BTR's - NEDO-2.4154, October 1978 (ODYN).

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