ML19277C596

Forwards marked-up Revised Draft FSAR Page Changes Per Draft Ser,Resolving Discrepancies & Incorporating Gestar II & Odyn Analyses of Pressurization Transients Into FSAR

ML19277C596
Person / Time
Site:	Limerick
Issue date:	07/12/1983
From:	Bradley E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Schwencer A Office of Nuclear Reactor Regulation
References
NUDOCS 8307150069
Download: ML19277C596 (7)

Text

PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX BC99 PHILADELPHI A. PA.19101 EDW AR D G. B AU ER. J R.

(215)841-4000 2'::::::,*:" ......

EUG ENE J. BR ADLEY a ssoceave e..anak couesseh DON ALD BLANKEN RUDOLPH A. CHILLEMI E. C. KI R K H A LL T. H. M AHER CORNELL PAUL AUERB ACH a.....a ... . m c....., y ,

EDW A RD J. CULLEN. J R.

THOM AS H. MILLF9. J R.

IR EN E A. Mc K EN h A .

assestasuv cou-sa6 Mr. A. Schwencer, Chief Licensing Branch No. 2 Division o f Licensing U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Subject:

Limerick Generating Station, Units 1 6 2 Core Performance Branch and Reactor System Branch Draft Safety Evaluation Report Re ference : 1) A. Schwencer to E. G. Bauer, Jr.

letter dated March II, 1983

2) E. J. Bradley to A. Schwencer, letter dated May 25, 1983

3) E. J. E'.dley t o A. Schwencer, letter dated June 13, 1983 File: GOVT l-1 (NRC)

Dear Mr. Schwencer:

In the process of incorporating GESTAR II and the ODYN Analysis of Pressurization Transients into the FSAR, reference letters 2 and 3 respectively, it was discovered that editorial changes to FSAR section 5.2 were required. Th! attached revised draft FSAR page changes resolve those discrepar.cies where GESTAR II and ODYN page changes were inconsistent. The information contained on these draft FSAR page changes will be incorporated into the FSAR, exactly as it appears on the attachments, in the revision scheduled for July, 1983.

Sincerely, 8307150069 830712 PDR ADOCK 05000352 E PDR RJS/cw/P-98 Euge.e J. Bradley @\

7 3 cc: See Attached Service List I

i

\

cc: Judge Lawrence Brenner (w/o enclosure)

Judge Richard F. Cole (w/o enclosure)

Judge Peter A. Morris (w/o enclosure)

Troy B. Conner, Jr., Esq. (w/o enclosure)

Ann P. Hodgdon (w/o enclosure)

Mr. Frank R. Romano (w/o enclosure)

Mr. Robert L. Anthony (w/o et. closure)

Mr. Marvin I. Lewis (w/o enclosure)

Judith A. Dorsey, Esq. (w/o enclosure)

Charles W. Elliott, Esq. (w/o enclosure)

Jacqueline I Ruttenberg (w/o enclosure)

Thomas Y. Au, Esq. (w/o enclosure)

Mr. Thomas Gerusky (w/o enclosure)

Director, Pennsylvania Emergency Management Agency (w/o enclosure)

Mr. Steven P. Hershey (w/o enclosure)

Donald S. Bronstein, Esq. (w/o enclosure)

Mr. Joseph H. White, III (w/o enclosure)

David Wersan, Esq. (w/o enclosure)

Robert J. Sugarman, Esq. (w/o enclosure)

Martha W. Bush, Esq. (w/o enclosure)

Spence W. Perry, Esq. (w/o enclosure)

Atomic Safe;y and Licensing Appeal Board (w/o enclosure)

Atomic Sah ty <ad Licensing Board Panel (w/o enclosure)

Docket and.Se;vice Section (w/o enclosure)

-m LGS FSAR i u D; 'qh

% er

a. Prevent overpressurization of the nuclear system that could lead to the failure of the RCPB

b. Provide automatic depressurization for sv.all breaks in the nuclear system occurring with misoperation of the high pressure coolant injection (HPCI) system so that the low pressure coolant injection (LPCI) and the core spray (CS) systems can operate to protect the fuel

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barrier -

c. Permit verification of its operability

d. Withstand adverse combinations of loadings and forces resulting from normal, upset, emergency, or faulted conditions -

5.2.2.1.2 Power Generation Design Bases The nuclear pressure relief system MSRVs are designed to meet the following power generation bases:

a. Discharge to the containment suppression pool

b. Correctly reclose following operation so that maximur operational continuity can be obtain %;4;m, og dt,w,n f *o 5.2.2.1.3 Discussion

_f Ret. 5 2-s).

The ASME B&PV Code requires that each vessel designed to mee Section III be protected from overpressure under upset loydi ivds. Ti.e code aA20 = a peaA aAAowco2e pressure v! l a v >. oI) veasel des'. pressure -

r upse nditions The co' specific . ion for sa'.ty lve reg ire th th low st afe y va.ve b set a or low ve sel desig pre sure d hat e hig ac safety s . e be set that to- accumula d pressure n,o_t arcead l'ai cf the decigr. preocure for upcet rcrfi+ inns.'

nes setpoints satisfy the ASME Code specifications for safety valves, because all valves open at less than the nuclear system hc design pressure of 1250 psig.

h3b The automatic depressurization capability of the nuc' tear pressure relief system is evaluated in Faction 6.3 and 7.3.

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The following detailed criter  ? used in the selection of MSRVs:

a. Meet the requirements of ASME Code,Section III

b. Qualify for 1005 of nameplate capacity credit. for the overpressure protection function k

5.2-2

sic k~

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c. Meet other performance requirements such as response time, etc, as necessary to provide relief functions The MSRV discharge piping is designed, installed, and tested in accordance with the ASME Code,Section III, Class I.

5.2.2.1.4 Main Steam Safety / Relief Valve Capacity The MSRV capacity is adequate to l'imit the primary system pressure, including transients, to the requirements of the ASME B&PV Code,Section III, Nuclear Vessels, up to and including the Summer 1969 Addenda for Limerick. The essential ASME requirements that are all met by this analysis are as follows.

It is recognized that the protection of vessels in a nuclear power plant is dependent on many protective systems to relieve or terminate pressure transients. Installation of pressure-relieving devices may not independently provide complete protection. The MSRV sizing evaluation assumes credit for operation of the reactor protection system, which may be tripped by either of two sources: a direct or flux trip signal. The direct scram trip signal is derived from position switches mounted on the HSIVs 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 prior to 10% travel of full stroke. The pressure switches are actuated when a fast closure of the turbine control valves is initiated. Credit is taken for the safety / relief valves in their ASME Code-qualified self-actuating mode.

The rated capacity of the pressure-relieving devices is sufficient to prevent a rise in pressure within the protected vessel of more than 110% of the design pressure (1.10 x 1250 psig

= 1375 psig) 2.

for events defined in Secti%stch)

E Full account is taken of the cressur ro on both the inlet and di_scharoe sides of the valves'. 'l MSRV) discharge into the IC g434 -suppression pool through a6 discharge pipe from each valve that is designed to achieve sonic Tlow conditions through the valve, thus providing flow independence to discharge piping losses.

Table 5.2-5 lists the systems that could initiate during the design basis overpressure event. .

5.2.2.2 Desian Evaluation 5.2.2.2.1 Method of Analysis

.The nuclear b ilu s fem peuv re Pro fedl* WAJ d t.Sipied VJin3 Tc facion tha peace" a f er._ t h : "uricer-bM lar svstsew, r--+6-tin-(- extensive analytical models representing all essential dynamic

' characteristics of the system cr e-1 2 t;d on a large computing 5.2-3

1 9 g4tre nce, E. 2.~ 2 a -for -the fo;nt kinett'es mzodel (MDv') an d D.eS e rence 6.'2.-26 for fhe one- dienensten/ hine he s modaj ( OOygL LGSFSAR ~ rm t n..% -

Mte 1 Cr 'N 2 h.!b-: a 3(-

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 principal controller features, such as feedwater flow, recirculation flow, reactor water level, pressure, and load demand. These are represented with all their principal nonlinear features in models that have evolved through extensive experience and favorable comparison of analysis with actual BWR test data. ,

mode 18 b ocumented in b ; n;.ng a/ etailed description 3cf

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rf T: I-t Tr: r ic ..

MSRVs are simulated in A P :: :t.: c a nonlinear representation, and the model thereby allows full investigation of the various valve response times, valve capacities, and actuation setpoints that are available in applicable hardware systems. ,, .w o ,, , ,4 ihen meddr are

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"arteristics are also sirulated in the modelCn GESrtWll ([ld U-0.

o 5.2.2.2.2 System Design A pararetric study was conducted to determine the required steam flow capacity of the MSRVs based on the following assurptions.

5.2.2.2.2.1 Operating Conditions

a. Operating power = 3435 MWt (104.3% of nuclear boiler rated power)

b. Vessel dome pressure P <1020 psig

c. Steamflow = 14.86 x 106 lb/hr (105% of nuclear boiler rated steamflow)

These conditions are the most severe because At lowermaximum stored the power conditions energy exists at these conditions.

transients would be le~ss severe.

5.2.2.2.2.2 Transients The overpressure protection system must accommodate the most severe pressurization transient. There are two major transients, the closure of all MSIVs and a turbine-generator trip with a coincident closure of the turbine steam bypass system valves, 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 indirectly derived scrams; therefore, it is (

used as the overpressure protection basis event,and h - 2:7 5.2-4

a y p't.s p g

< LGS FSAR sikd ub E

' 51 g' e5 1. ab d.e-e 42das tne sequence oi eve ts ict v.i na n eam ne . ol ' o .

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t yL e i neie l l e d McCV cacacity 5.2.2.2.2.3 Scram

a. Scram reactivity curve - Figure 5.2-3 [or REDy and

'Flsun G 2-Ib 4e- 00Vu Control rod drive scram motion Figure 5.2-3 b.

5.2.2.2.2.4 MSRV Transient Analysis Specifications

a. Valve groups: 3 .

b. Pressure setpoint (maximum safety limit):

1. 1142 psig - group 1

2. 1152 psig - group 2

3. 1162 psig - group 3 The setpoints are assumed at a conservatively high level above the nominal setpoints as shown by Table 5.2-2. This is to account for initial setpoint errors and any instrument setpoint drift that might occur during operation. Typically the assumed setpoints in the analysis are 15.Tto 2%)above the actual nominal setpoints. Highly conservative MSRV response characteristics are also assumed. gg-5.2.2.2.2.5 MSRV Capacity Sizing of MSRV 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 (Section 5.2.2.2.2.2).

5.2.2.2.3 Evaluation of Results 5.2.2.2.3.1 MSRV Capacity The required MSRV capacity is determined by analyzing the pressure rise from an MSIV closure with flux scram transient.

The plant is assumed to be operating at the turbine-generator design conditions at a maximum vessel dome pressure of 1020 psig.

The analysis hypothetically assumes the f a: lure of the direct isolation valve position scram. The reactor is shut down by the backup, indirect, high neutron flux scram. For the analysis, the saf ety setnoints are assumed to be in the range of 1142 to 1162

( psig. bWTiD indicateK that the design valve capacity is

( capabl or maintaining an adequate margi,1 below the peak ASME Code llowable pressure in the nuclear system (1375 psig).

Soih N R.ED/ a.J  %. ObyAl an{Q b.2-3

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Figure 5.2-1a. C curves produced by ttds analysis. The sequence of events in Table 5.2-Etassured in this analysis was investicated to meet code requirements ar.d evaluate the pressure relief system exclusively.(The sessence e t evedr as.ro,ord and the corves het4 6y th e ODWanalysic arc Pmen4c4 in Gl,h. 5.2 y y

. er the General Requirements Ior erotection Against Overpressure as given in Section III of the ASME B&PV Code, Ft3on 5.2,. j g credit can be al. lowed for a scram from the reactor protection In addition, credit is also tak'en for the protective re rpchedD system.

circuits that are indirectly derived when determining the required MSRV capacity. The backup reactor high neutron flux scram is conservatively applied as a design basis in determining the required capacity of the MSRVs. Application of the direct position scrams in the design basis could be used, since they qualify as acceptable pressure protection devices when determining the required safety / relief valve capacity of nuclear --

vessels under the provisions of the ASME Code. [This ftpre shws Thb L COYt] Pre dMr keyer pean

• The parametric relationship between peak vessel (bottom, pres stef W, y

,and MSRV capacity for the MSIV transient with high flux and y position trip scram is described in Figure 5.2-4.g Also snown in Figure 5.2-4 is the parametric relationship between peak vessel (bottom) pressure and MSRV capacity for the turbine trip with a coincident closure of the turbine bypass valves and direct scram, which is the most severe transient when direct scram is 7 considered. Pressures shown for flux scram result only with multiple failure in the redundant direct scram system.

The time response of the vessel pressure to the MSIV transient with flux scram and the turbine trip with a coincident closure of the turbine bypass valves and direct scram for 14 valves is 111ustrated in Figure 5.2-5. Onlo shows that the pressure at the vessel bottom exceeds 1250 psig for less than 5.8 seconds, which is not long enough to transfer any appreciable amount of heat into the vessel metal and which is at a temperature well below 5 5 0 0 F_a_t the ctart of the transient.

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5.2.2.e.a.. Fu usaa t prop in inlet and Discharge Pressure drop on the piping from the reactor vessel to the MSRV is taken into account in calculating the maximum vessel pressures. Pressure drop in the discharge piping to the suppression pool is limited by proper discharge line sizing to prevent the backpressure on each MSRV from exceeding 40% of the valve inlet pressure, thus ensuring choked flow in the valve orifice and no reduction of valve capacity due to the discharca piping. Each MSRV has its own separate discharge line.

kr I3 hi$* rt show thd the Jettel fff!1**% prid & Qb & N *b N,i frei.

t (lQgf and O Dff) are considerdif lower Ihm fHe ASfW Codt xllwak of 1370 p,.t:3 F h -c (+

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