ML19329B640

Submits Reactor Sys Branch Evaluation of Facility Ser.Addl Info Requested.Applicant Commitment Required in Several Facility Design Areas

ML19329B640
Person / Time
Site:	Davis Besse
Issue date:	08/28/1975
From:	Stello V Office of Nuclear Reactor Regulation
To:	Moore V Office of Nuclear Reactor Regulation
References
NUDOCS 8002050755
Download: ML19329B640 (34)

Text

.-

+

O OO File D

D NRR Reading File CD RSB Reading File

~

o D ~9~I A

h)ocketNo.50-346

- "5-)

dL O

C Voss A. Moore, Jr., Assistant Director for L W s, Group 2, RL DAVIS-BESSE 1 SER INFL"T Plant Name: Davis-Besse 1 Docket No.: 50-346 Milestone No.: 01-24 Licensing Stage: OL Responsible Branch and Project Manager: L'#R 2-3. L. Engle Technical Review Branch Involved: Reactor Systems Branch Description of Review: SER Input Raquested Completion Date: August 15, 1975 Review Status: Complete The enclosed report contains the evaluation performed by the Reactor Systems Branch on Davis-Besse 1.

Reactor Systems review included Sections 1.5, 4.1, 4.4, 5.1, 5.3.2, 5.3, 5.5, 6.3.1, 6.3.2, 5.3.3, 6.3.4 and Chapter 15.0 from the Standard Format.

Revision 1.

There remains several areas of the Davis-Besse i design which require cc:mnit:nent by the applicant:

1.

Upgrading the valves in the croseaver lines between the LPIS and HPIS from hand-operated to motor-operated.

2.

Upgrading the crossover line betwen each LPIS train fresa active components to passive components.

3.

Adding the capability for leak-testing the check valves ia the hv ;rassure-high pressura LPIS interface.

l l

The applicant was officially informed of the first two iteam by letter on December 26, 1973. Our position on the third item was i:

transmitted by letter to the applicant on April 18, 1975. Also, 4

we require additional information in several areas which could p

possibly lead to design changes. These areas are further specified in the enclosed text. In addition, see Section 1.6 for a more I

detailed summary of a nseber of technical and administrative changes.

Orieinal Signed by Victor Std!o Victor Stallo, Jr., Assistant Director for Reactor Safety

^ i.iu2 vi M

~ Revi w v

o**i==*

S

_TR SB _.

_TR:RSB

_ TR J,/J.

Enslosures Davis-Besse 1 Evaluation-W

/*.

RBaer_ ___

.ak_

.u-'"*

-.B /f D5._ BL )$ J15..._ _8L/D"75 8L.

.. ac.u.a

.,.,n ~ ~

W OM 4

e -*

e

\\

AUG e 0 Ib/$

cc:

S. Hanauer

..g. minamen

n. nord

w. Mcdonald A. Schweneer L. Engle V. Stallo T. Novak R. Bear G. Hasatis (5)

~

0 9

sum 4AMt >

,_ _,,, _.,,,, _.m

.._.a,_..

_. a _.. --

._a... _ ___

• 1. 5 liequirements for rurther Technical Infernation The applicant has identified in Section 1.5 of the Davis-Besse 1 FSAR development progra s applicable te the Davis-Besse 1 design.

These programs were initiated to establish the final design and have each been completed.

'n'e conclude that the research and devel-opment test prograus outlined in the Davis-Besse 1 FSAR provide the infor=ation necessary for the design and safe operation of Davis-Besse 1.

t i

t i

i i

I

1.6 Facil.ity Modifications as a Result of Regulatory ":a f f Review During the review of the Davis-Besse 1 operating license application, the applicant proposed or we requested a number of technical and administrative changes. These changes are described in various amendnents to the original ar. plication. Ue have listed bc1cu the more significent redifications that have been or vill be required to be made as a result of our review. The sections of this report where these =atters are discussed core fully are noted in parenthasic.

1.

Upgrading the valvec in the crossover lines between the 12I3 and HTIS from hand-operated to motor-operated (6.3.2).

2.

Upgrading the crossover line between each L?IS train from activecomponents(notor-cperatedvalvec)topassivecomponcats (cavitating venturi) (6.3.2).

3.

Addition of testing corponents for tha low pressurc-high pressure LPIS intcrfc - 4c'cek :1 c CT 10f".1

d EP 70/.-~) (5.5.7).

4.

Addition of preoperation tests to derenstrate cperability of i:he local can":.1 Sne..:cul backup en ecc'.: CCCS v.sive, and to denenstrata the capability of the ECCS to operate in the recirculation tode (6.3.4).

5.

Changes in the Davis-3csce 1 Technicci specifications to prohibit all partial loop operation (4.4).

6.

.', ' ' '.r !. u o f

.a ce re. f i c. t ;-

lty ir. ', n r. ;]--: :-

m'ic de A; a of tht: RL3 nid cer to acco"nt for a poteatini.ly stuck open vcat valvc(4.J).

D D

CV CJ rJ rm.

D g

y

_A J

~

s s

7.

Addition of a periodic surveillance requirecent in Technical.

Specifications for venting of all ECCS lines and pu=p casings to minimize the potential for a trater ha= tar (6.3.4).

4 e

s

,3 i

, 4.0 REACTOR 4.1 Su. mary Description The design of the D&W reactor for Davis-Besse 1 is cimilar to the design of other pressuri:cd water reactors that we have recently 1

approved for operation.

The cora consists of 177 fuel asse=blies i

having 203 fuel rods each; the design heat output of the core is 2772 }Wt, which is the same as the design output for the Rancho Seco Full and part lorsth control rods, dissolved boron, and burnable core.

poison rod assemblies (BPRA) are used for reactivity centrol.

A unique feature of the Bi'.J design is internal vent valves 1:hich minimize steam binding in the event of loss-of-coolant accident (LCCA).

i

.The prinar/ difference bet ee:t Davis-Basse 1 and Usacho Seco is the raised steam generators in Davis-Ecsse. These higher stecs ge.terators 4

further decrease the potential for steam binding in the event of a LOCA.

i 4.4 Thermal and Hydraulic __Desi3n The Davis-Ecsse 1 reactor is designed to cperate at core power

.levela of up to 2172';i..*t. 'faica ccrees;c.id.; to a nat electrical output of about 906 >T.fe.

Ua have evaluated tha thermal hydraulics on the b: sis of' 27721".it.

Davis-3aase 1 will utilize : 15x13 fuel assembly as in the Rancho Soco plant. _As shown in Table 4.4-1, the thermal and hydraulic design paraceters for the two plants are similar.

The principal criterion for the thermal-hydraulic design of a reactor is to prevent fuel rod da:.tage by providing adequate heat ti.mier for h6 e.-rim:c cu Imat sn. rat.ica 7atrTrnc ecuurric.

P D

'DJ-1 v

L em

~ T D

v JU s.I.

_a

~..

~. ~

.._ - =. -.-. -

. = _

t g,

'm

. - j dt$ringnormaloperationandanticipatedoperationaltransients.

Maintenance of nuc1cate boiling is a basic objective of a thermal-hydraulic design. The applicant has demonstrated, through the use of the Westinghouse W-3 correlation, that a departure from nucicate boiling heat flu:: ratio '(D:;3R) greater than 1.30 is maintained for steady stato and anticipated trcnsient conditions.

We have required that the app *1 cant consider the effect of a p

stuck open vent valve on the analyses of the charcal-hydraulic design 1

of the reactor ecolant system and core and for all transients. Before power oparation, the applicant must either; (1) submit the reanalysee, (2) show that a stuch open vent valve would be detected by an operator, i

or (3) show that vaat valves are not stickins open on o:ecating recators.

4 Inasmuch as the applicant has not presented information regarding 2g i

j items (2) and (3), the staff requires that one valve less than the 4

ninimum detectabic number of stuck open vent valves be assumed open and a

the corresponding core flou penalty be impocod for the thermal-hydraulic design of the RCS and corn.

The applicant la required to provide thi.c i

analysis which will be evaluated to determine the =a::imum pcuer level 3

of the sy.dte.

Further, we vill Zeo r? quire rhnt th: vcat valven I:c tested during each refueling.

Another parameter that influences the tbermal-hydraulic design of 1

the core is rod-to-rod bouing within fuel assemblies.

The applicant id analytically predicted-the amount of bouing ubich could occur at the elr.Jdin: hot spot durir nn operat.ica il tre. orient ulc5 v. em;t: 0?

eulit ion.

Th. an :1ys ie..cac pe ct'ar.;cd at. 100%,wucr eith an assuned 1

v4.or:.il fle.t eendith u (lulet flow blochage).that ucuid c ase N;'. s ppm D

D l a s 1<

o

,D1 k S.1

._ 3

-G.

Operation at checc postulated conditions resulted in a calculated maximum fuel rod cladding temperature of 1025 F and bowing of 49 mils.

During the Oconec 1 refueling, six fuel assemblies were examined visually and df=ensionally.

? ster channel and line scan ceasure=ents indicated a maximum rod bou of approximately 30 mils. 30W feels i

that the obccrved rod bcu is acconnedated within the current design approaches and is pursuins a progran to demonstrate this.

BSU generically plans to develop bow correlations and predictive techniques to analyse the data and the predicted bou from a thernal-hydraulic standpoint.

Tha staff intends to follou this program and will considtr the application of our conclusions to Davis-D2sse 1.

Since the applicant does not propose to validated operation of the plant in a singic loop configuration (i.e., two pumps in one loop running while both pumps in the other loop are idle), the Technical Specifications will prohibit single loop operation. Also, the applicant has been requested to further support other partial loop configurr. tion". by prs it u c3 a LCCA snal ais duri..; ej:.rctRn in

'.Mr.

j

= ode.

Until this analysis has been revicued by the staff, Technical SpeciJications vill not allow partial iccp operation.

On the basis of our review of the thermal-hydraulic character-1 istics of Davis-Besse 1, including a compar'ison with the previously approved Rancho Seco, uc conclude that with the stipulations noted above, the thermal-hydraulic design of Davis-Besse 1 is acceptchic.

D cm em m)

D JL o

g7 g

_S__tLn.1_a

.s i

1 l

i

T Table 4.4-1 Therms 1-Hvdraulie Desinn Sutrary Comoarison of Davis-Besse 1 and Rancho Seco Davis Rancho Besse Seco Design Core IIcat output,IfIt 2772 2772 Nominal System Pressure, psia 2200 2200 I

Vessel Coolant Inlet Temperature, 'F 555.4 557 Vessel coolant Outlet Temperaturc 'F 608.6, 607.7 Total Heat Tr.;nsfer Surface Area 49,734 49,734 in Cera, ft 9

Average Heat Flux, Btu /h-ft-185,090 185,090 4

9 Maximum Heat Flux, 3tu/h-ft-554,200 576,835 Average Thermal Output, kU/ft 6.105 6.105 j

!!aximus Desi;;n Thermal Output, kW/ft 18.28 39. 0 3 Maximum Cladding Surface Temperature, *F 654 654 Averaga Core Fuel Temperature, *F 1200 1200 I'c :

• r m "u ci Tc.~.p sro tu t e e t.:: : ? pet, *F 1050 4170 Total Reactor Coolant Flet., 10 lb/h 131.32 137.0 Core Average Coolant Velocity, fps l 15.74 15. 5 2

. DNB Ratio at Design 0verpo::ar 1.41 1.39 DNB Ratio at Design Power 1.79 L75 F1]

ILUL

m.

s

_ S.

e 5.0 REACTOR C001 i:1T SYSTD1 5.1 Summarv Descri, tion Davis-Bense 1 uses a B&W 2-loop nuclear steam supply system.

In most important aspe ts, it is the same as the Rancho Seco system.

The pri=ary difference is the higher stens generators on Davis-Besse provided to decrease the potential for steam binding following a LOCA. On the basis of our evaluation of the Davis-Besse system and the similarity to the previously approved Rancho Seco, we conclude that the overall design of the reactor coolant system of Davis-Bosse 1 is acceptable.-

5.2.2 Overpressure Frotection Overpressure protection in accordance with the AS E Loiler and Pressura Vessel Code,Section III, Article 9 is provided by pressure relief of the KCS from two pressuriser code safety valves and one electrically actuated relief valve mounted on no::les en the pres-suriser. The valves discharge through ncnifolding to a pressurizer quench tank. The code soiety valves are each rated to carry 336,000 lba/hr at 2450 psig, which is the maximum calculated surge of the systa. L.utd upen the nors:. presaure rra nient.

Tha electre. atie rc. lid valve has a capacity of 100,000 lbm/hr at 2255 psis.

The pressuricer safety valves are sized on the basis of'the. cost severe pressure transient i= posed on the RCS. The applicant's cualyses of safety valve capacity (bah'-10043) show that the upsets that produce the lar;;est pressure transients are the co: trol' red withdrawal fr.n 1:n. pcNer end the turbine tria fri thi o.crpc.ter ceu,itian.

.10 v;h uc at:a f reviez c2 h.Y.:-icuu !:.~

not e

mi U. ca ce? L*tcq, e e.n S i n;* d 0 d t h,:

3.'" :!daly t i. cal, -D.

*; S 0 'Jt.tr 3t-Iijcl 1.

gas orD'9 3h 1],1.M 5

.a 1

y

, + - -

---m 4-

y_

s.

9 margins e:r.ist to conclude that the Davis-Besse 1 design is acceptable.

5.3 Themal-Hvdr,ulic Svstem Desin The ther.a1 and hydraulic desiga bases of the RCS are discussed in Section 4.4.

~ - -

- - u -w.

-... - ~ ~ _ -

5.5 connenent and subsystem Decien 5.5.1 Reactor coolant Purns and "otors The Reactor Coolant Pu=p is designed to provide adequate core I

i cooling flou and hence sufficient heat transfer to maintain a D:!LR 1.30, within the para =eters of operation.

Sufficient pump rotational inertia is provided by the flywheel to j

r the provide continued flow folleving a 1 css of pump pmrer such that reactor neutroa power can be reduced before D:'E limits are exceeded.

1 J

5.5.2 Steca Generator The steam generator is a vertical straight-tube-and-shell heat exchanger and produces superheated steam at conocent turbine throttle pressura cvar the operating pcVer range. The primary reactor ecolant enters the steam generator upper he=f spherical head, flows downward i

J inside the tubes giving up heat to generate s:cca on the shell side

] -

secondary loop.

'The tube and' tuba-sheet beundcry have the cano design pressure i-and tc:.peraturc as the r:.acter c h..a synten.

Siac:. c::c :! c u generators must provide a heat sink for the pricary reactor coolant

' sys tc.a. they are at a higlior clavation :han tha core to sesure 1.'

natural circulation for decay heat renoval.

5.5.3 neactor coolant Piping The reactor coolant piping is designed and fabricated to accommodate

-the systen r. cssm es a:d terperatures attained under all expected codes

.. m:. 2.w e. :lo.u.

of plant oper:n ion er cnw p.ned D*

1l ov s

DJ A

L v1 S_i

_a n

-__w-9 5.3.4 Main steamline Flow Restrictors The applicant stated that because of the small inventory of water in the gSW OTSG design, no flev restrictors are required in the main steamline.

This contcution is supported by their analysis of postulated stean line breaks (FSAR Section 15.4.4).

5.5.7 Decav Iteat Removal System The Decay ' dent Removal System is designed to remove decay heat and sensible heat from the RCS and core during the latter stages of cooldown. The system also provides cu::iliary spray to the pressuri:cr for complete depreasurization, maintains the reactor coolant temperature during refueling, and provides the means for filling and drainin; the refu211 ; cavity.

In the event of a LOCA, the decay heat removal pumps are used for low pressure injection of borated water into the reactor vessel for e crgency core cooling.

The Decay lizat Removal Systen is placed into operation approninately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after initiation of plant shutdoun when the tc perature and pressure of the IsC5 are cold.' 230'F cnd 260 psi;, respec:ively.

Assuning that tuo pumps and ecclers are in service, cnd that cach cooler is supplied with compencut coeling vater at design flou and temperature, the DHRS is designed to re, duce the RCS temperature to 140*F within 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.

If one of the two pumps or one of the two ecolers 'is not operable, safe cooldown of the plant is not coeprecised; however, the time required for cooldevn is c:: tended.

The applicant has sheun that, assumine c-D fi l @ M

&&JC m sy n

WDs _1k

.-a.-

-- - - --- u - - --

s one train is available,.the plant can be shut down to below 212'F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. To increase the reliability of the DERS, I

the applicant has installed a manual bypass in the DHRS suction line.

The applicant has sho:in that, should motor-operated suctica valves DH 11 or Dl! 12 be discovered failed closed at the time shutdown cooling was needed, the operator is able to enter the containment and open the manual bypass valves without e:cceeding dose limits. This radiological assessment is under review by the staff.

In addition, the applicant is required to show that, should a spurious closure of DH 11 or D11 12 occur during RHR system operation, either sufficient tiac cxists for the operator to detect the less of flow and secure the icw pressure pumps before overheating occurs, or the existing design is able to i

cope with such a loss of flow until the manual bypass has been opened.

The DIES design for Davis-Besse 1 has double isolation valves cn the suction side to isolate icw pressure components from the reactor ecolcat sfatem. The staff requires that the applicant ensure that the design t

features which protect the DER sy tem against overpressuricetion during l

ahutdewt. 6.Y.la the w t/cten : clicy tyaten u fu: cden n ) cra ade.rc:.c.

a The applicant is required to provide analyces :hich justify the DI!RS relief capacity. These analyses are to consider the occurrence of a,-

worst-case pressure event under these shutdown conditions.

D D

1 cw a'g m

Y 3

N.]

.}L 5

a

,.v I

s n

m......_.

-.%__~._.<...-

. ~ _ - -. _ _ - -

s 4

1-

/

i 5 5.10 Pressurizer The pressurizer =aintains the RCS pressure during steady state operation and l'inits precsure change.s during transients.

It contains a water voluna, sized to provide the ability of the systc= to experience a reactor trip and not uncover the lev level sensors in the bottom head and to naintain the pressure high enough so as not to activate the high pressure injection systc=; and a volu=c of seca=, sized to provide the ability of the syste.2 to experience a turbine trip and not cover the 1cvel sensor in the upper neal.

i

l Electric hea:cr bundics, located in the icver section, and a water spray no:::lc in the upper sectica maintain the. creau and.rster at the saturatien temperature which corresponds to the desired rec ter coole,nc syste;::

i During outsurges, as the RCS pressure decrenses, so:s of the pressure.

water f L;shes to steca and the electric heaters rectore the norcal operating pressure, During insurges, as RCS pressure increcses, the veter spray cendenses s ca= to reduce the pressur...c the nornal operating lav al.

Tuo ASE cede screty valves are connected to the upper pressurizer head to rel:..va c:m te; ever7:cc.;ura.

A pilet-:perc;cd rcJir.f.:1ve is 2152 provided to-limit the lifting,. frequency of h2 code safety valves. The 4

f safety and relief valves discharge to tha..ressuricer quench

tank, i

located uithin cont:in=cnt.

V e J{L D

mpy.

o il

_ 3.

L a, -

l l

-m.c____...

m___

1 4

5.5.13 Safety and nelief valves The pressuriser safety valves are bellows sealed, balanced, 2

spring-loaded safety valves which are provided with a supplemental The backpressure balancing piston for handling a bellows failure.

pressuricer relief valve is an electrically actuated, electrically controlled, pilot operated, pressure loaded, relief valve.

The combined capacity of the pressurizer safety valves is 672,000 lb:a/hr, which uas bcsed on twice the maximum surge resulting froci the upset that produces the largest pressure transient (see su*c-section 5.5.2).

The maxinue surge assur::es no direct reactor trip, operator action or credit for actuation of the pressuriser relief valve or turbine bypass systca.

The pressuriser safety valves J

prevent the reactor coolant system pressure from exceeding 110% of system design pressure. The pressurizer pouer operated relief valve prevents un:lesirable. lifting of the spring-loaded safety valves.

l 5.5.14 Internhls Vent Valves The care s t: Tort vent valvos are located on a cocmon plane in These the upper cora support wcldsent above the outlet noz:lca.

valvesprovideadirectflowpathbetweengheuppercoreregion i

and inlot annulus in the event of a' loss-of-coolant accident from an inlet line break. This flow path provides for pressure equali-nation by the ventin7, of steem to the bresh and per=its the emergency e M :., r.- v r :., r.. ; md tn r ie n.

.: m :re ic tr vent c-:

each t.ith :.n circeriva flev d L:::nrer of 14 inches. The fere of ti:e J b:e is slight ly inelir.ed to inaure a positive seal of the _:

%t he D

D v

0 m=

- _. ~

s differential pressure across the valve. The individual vent valve design is crsentially the same as on the Oconee Class plant.

In the thermal-hydraelic analysis of Davis-Ecsse 1 for norcal cperation, the applicant assunad that there,tas no core bypass flow resulting fron an opsn vt.nt valve. At present, there is not adequate instrum.cntation to detect the system flew change (approxicately 57.

reducrien in core flo't) uhich.could result fron an open valve.

The staff position has not changed from that taken en the Oconee plant.

Further, the applicant has not prescated data frc= operating plants to shou that a stuck cpen vent valve is an c:-:tren21y lo.: probability event. Therefore, ue aill require the ther al-hydraulic reanalysis described in Section 4.4 We conclude, subject to the conditions as noted above, that the proposed reacter ecolant syste=, subsystens, and cor.ponent designs are acceptabic.

4 P #

]D D

e.s o Ju o

D

@ l A

.S.

.:s a

_-_._m

_4

+

5.5.15-Loose Parts Monitorin: Svstem Occasionally, misec11ancous items such as nuts, bolts, and other'small itens have becc=c loose parts within reactor coolant systems. In addition to cau::ing operational inconvenience, such loose parts can da= age other cceponents uithin the system or be an indication of unduo wear or vibration. For such reasons, the staff has encouraged applicants over the past several years to support programs dasigned to develop an effective, on-lino loose parta conitoring system.

For the past feu years us have required many applicants to initiate a program, er to parckipate in an ongoing program, the ohjactive c? tchich uas the devcicprant of a functional, Icose parts monitoring system trithin a reasonable period of time. Recently, prototype loose parts =enitoring systems hava been developed and are presently in operation or being install'ed at several plancs.

Such a system has been in-stalled on Davis-Eesse 1 to provide the opcrator iith an audible and visual alarm of icose parta which accuaulate in the bottcm of thc. react.v vo:s;1 : na : ::a.t:--

2:~ cec > m:- ; cue::: 1r.

staff vill follou the performanc2 of this on-line conitoring

. system.

e

)

$ $10 Wo g,7 u O-A ie

... ~ -....,

m.

.2 a

-~~.

s -

'N 17-6.3 EmerSency Core Cooline, System (ECCS) 6.3.1 Desien Bases 4

Toledo Edison Company has stated that the Davis-Besse 1 Escrgency Core Coolin3 Systent will be designcd to provide core cooling during postulated accident conditions uhich occur when mechanical failure in the reactor coolant system i

piping results in a loss of coolant from the reactor vessel greater than the I

i available coolant takcep capacity using normal operating equipment. The ECCS equipment is desigaad to provide both short and long-term core cooling l

capability.

The applicant's desi;:n bases are to cnnure that the core uill be i

cooled and vill not lose its geometric cenfiguration by torninntint; tha to:::perature transient for any siac break up to and including a double-ended rupture of the largest primary coolant line.

The applicant states that these requircaents uill be cet even with mininum engineered safeguards 1

available, such as the less of one czersency power bus, together with the i

l loss of offsite pcuer.

The ECCS to be provided is stated to be of such number, diversity, relienility aad raducc.:.u/ :ha: no c.in,12 Bij ure c:. l.003.:quy..'ac ccearring durir; ;

LCCA will result in inadequate cooling of the reactor core.

Each of the ECCS subsystems are to be designed to' function over a specific range of

~

reactor coolant piping systcc break sines, up to and including the flou aran associatad with a postu}ated double-ended break in the largest reactor cNj a.l*J Ml'*2 (1/.. l. N 1;' the [nr;'VO*J N G tt b } t -* d r I.d "".* Cil).

O g

m i

D A

0 DJ U

V{j g V

p.3 D

l 9.y 1

vk S.k

_.l w.

-4, n

--g,

--a p

g-a,-

-e-,-

--e,-

---aw

+ - -- -, - - -. -

v

.c.---..

s t

6.3.2 Syctem Desinn The ECCS proposed for Davis-Besse 1 consists of core flooding tanks (CFT), a high preocura injection (IIPI) system and a low pressure injection (LPI) system.

Previsions are included for recirculation of the borated coolant after the horated water storage tank ('L'ST) is exhausted. Cc binations of these syste':'.s assure core ecoling for tha ce=plete range of postulated break sizes.

Followino, a postulat.ed LOCA, tha ECCS will operate initicily in the active high pressure injection code, the passive injection rode, then in t be activa icv presc.:ro injection : de, an:I sub-sequently in the recirculation mode, liigh pressura injection, upon actuation of an Engineered Safety Fcature Actuation Signal (ESFAS), will consist of the operation of two centrifugal li?I pumps (rated at 500 gpm each cr a design head of 2700 f t) thich inj et 1"00 p7= ccacentrated beric acid solution inte the reactor coolant system cold legr. These pumps take their suc ion fren rN b.rntad - :

7 t e r: g.: ta' uc a ' ' : a ce i t:.-.c

c. f 3'0,000, lle..s.

Low pressure injection will be accenplished through two separata lcw paths, each having one decay heat renoval pucp and cooler.

The icw prt.:::sure injection lines terminate directly in the reactor vessel th cu,h the core floodin~ no: les lece.ted in the reactor vescel.

"cr i.m rt -t e r cool i ;, the 2.n p e.m r2 injectic "i2 the A c ;y ::ea t re.

.il m :y irated

..t

_01_i,m

ne:. :t a d; ;L.

m; ' c f 35 : c :' :.,:

pro Jes 1:'00 y: t !.'o ro n o!.uen Ire a the bor. t c i ater st.'r:.,p tank. A crossover line cennecting the tv3D T!E t 'ithin OO o

D 9-

-oI-b

~

s the reactor building is provided so that if a single failure causes the loss of LPI flow in one path, part of the flcu frca the active LPI path can be injected into the reactor vessel through the piping associated with the inactive path. The cross-connect is also intended to accure abundant long-term cooling flow to the core in the event of a cort. ficodin;t line breah in addition to this single activa failure. Ue have require.d that this cross-connact be codified to inccrporate a passive network de-sign similar to that adopted on such plants as Arkansac Nucinar Gna Unit 1 and North Anna 3 and 4 (plus n.11 205 _ruel Assenbl/ plants).

This preferred nathod consists of crossev:r lines which contain no motor-operated valves.

Instead, this crossover net.crk utilizes the flow-lir.iting characterictic of a cavitating.enturi to prc, vide cn autcraatic split ?.a ECC water between the two LPI trains. The follouing cinplified din 3ran illustrates this principle:

/EfC.P:WP 3CCO f/2!??

.i /a.4 m

I >*

.:~.r *.'

3CM$?M

t

n-

~.:7.

.',f f

. : - r.+-y l

e-

-m

r ?

^ =

g i

e The obvious advantage of the latter method is that it does r.ot rely on-operator action to be initiatcd and it is less prone t: activa cenpenent failures.

Oom D

D<l Oh n

b v

.s a

s The ECCS will provide the long-term core cooling requircsonts by recirculating the spilled reactor coolant collected in the con-tainment sump back to the reactor vessel through the core flooding l

line nozzles. The changcover from low pressure injection to re-circulation is accomplished manually from the control room with automatic backup to the manual action.

For large sized pipe ruptures, the ECCS will provide the long-tcru cooling requirements by recirculating the spilled reactor coolant collcceed in the containment sump, bach to the reactor vessel via j

either of the available two trains of lou pressure pumps and coolers.

Prior to this time, the operator is required to shut off the HPI pumps to avoid their overheating uhen the S"ST valves are closed.

For ccall sized pipe ruptures, the reactor coolant system pressure may be higher than the maxi um low pressure injection pump head at the time containment sump water recirculation is required. Under this cir-cucatance, a cross-over connecticn is provided to parait alignment of the hih:t pic:sure esau-up pump suction al:h the lc.,ressure iajectica cc..._:

discharge to permit high pressure injection during the recirculation. ode of operatica. Presently, this alignient is ccc mplished by tha opa..:cc manually opening one valve in each of the two crossover pipe lines located i

s in the auxiliary building.

The staff requires that these valves be' cotar operated uith cow.tal and indicatio.t in the control room.

The pa.<uive Jojection mode of operation is provided by the core J.

o f r.:.

ch; eore :- Se c w:

1 Lu.

(1.:. 21 -

( f.':

amt la r.c

.,1 cd pipe b ru.:.. The coulant is auren.wienl.ly njectud.

.u.

thu I;CS prcmure drops bclou the core fi vd '-~' em mtank pressure (600 pci G.

D D<l JL n

D '9'T 0

r i

u o

- _ u,..

m..

.~..o. - - -.

4 1

~-

21.

f t

Each of the two core flooding tanks has a total volune of 1410 ft3 with 3

i a normal water volume of 10'0 f t' with 370 f t of nitrogen gas at a 4

normal operating pressure of 600 psig.

Each tank is connected by a core flooding line directly to a reactor vessel core flooding noz=le.

i The driving force for injection of the 1800 ppo borated water is supplied by pressurized nitrogen.

Each core flooding line will contain a motor-operated stop valve for isolation of the CFT during reduced pressure operation and two inline check valves in acries.

Since this i

portion of the ECCS involves a high pressure to low pressure interface, i

it is the staff's position that periodic checking of potential leahage 1

through chec'c valves CF 30/31 and DF. 76/77 is to be performsd nt locat r

annually.- This test is to be performed at or near normal reactor i

coolant operating pressure. The current design has a continuous monitor outboard of these two check valves, however, this location is not reliable in detecting the prelude to a pressure barrier I

failure (i.e., lankage of the inboard check valve).

To :. inly.1:a the pctentici for a >ater -ha=cr eacurring due to ECC water-being discharged into a dry lina, the t.ppliennt hcs,=tatef e'2:

during normal operatica, the ECCS lines will be maintained full by the static head created by the relative clevatigns of the SUST and ECCS piping.

I In addition, manual venting is provided at the ECCS pump casings and-i I

discharge piping high points.

The staff requires that the capability to'caintain filled ECCS piping be observed prior to startup cad that ebe V e lag er:nB ! :ns cc. n ;'.::..e a m. v.i h: r u n a i.11. c c. c i rc ---

a oT,T" L

1LUL

s

_ g, in the Davis-Besse 1 Technical Specifications.

Specifically, the section of 11PI piping inboard of normally closed isolation valves HP2A, HP23, lip 2C and llP2D cust be observed to be full cince during normal operation the static head of the BUST would be ter=insted at these valves.

O J oo1 93 r,

19 M a

.-...-..-..o_._.0.._..

s

- i Perfor. ance Cv.21ttar ten 6.3 3 Toledo Edison Ccanpny has stated that the emergency core cooling systems have been de.iinned to deliver fluid to the reactor coolant system to control thu predicted cladding temperature transient following a postulated pipe brea'e. and for renoving decay heat in the long-term, recirculation nodo.

On January 4,1974, Acceptance Criteria for ECCS uns published in 10 CFR Part 50.

The new ECCS criteria requires that:

(1) The calculated maxicam fu21 element cladding temperatura chall not exceed 2200*F.

(2) The calcul: tad total cnidation of the cladding shall norharc exceed 0.17 times the total cladding thickness before onidatien.

(3) The calculated total encunt of hydrogen genarated frc= the chanical reactien of tha cl:dlin;; uith water or steaa shall not exceed 0.01 times the hypotheticci ccount that would be generated if all of tha metal in the claddi:13 cylinder surrounding the fuel, encluding the cladding surrounding the plenum volute, were to react.

(')

'c: m I P ; ! c'. r n~,.c in :er a :,-

e

.1 : c.. 't

._ t..> c:r2 remains atend.ibic to coolin3 1 1 c ; c.ci a ef A : GC3, (3) l'ter na. c:ti.:.. c. ; d a =. a s. Z u. :

the calculated core tc perature shall bc raintained at an acceptably low valua nnd decar heat siiall be renoved for an cr. tended period of time required by the long-lived radioactivity ret..ainia; in the core, D o o m)

D fi l

oQ JL

':}Oc

- Q

? "- f.-

m F

a '. li. :. n t : aba;. t t.n

n can-

e.

4CS,.

v

-]

c,a.Iu if 21, 1E 75 GEfer nce 1) by referrace co asic;l report ;,X :- U... -

In addition to the rcrined LOCA ait.:iycis, the s taf f 's rev tcu of the I ' '

for D.. vin I sr.- 1 rennonend, <i. M ~ i...... t :., c,s... - t m, 4..

.o-

.. ~,." -...

~_

=-%

s

~

of minimum contain='ent pressure, single failure criterion, effects of boron precipitation en long term cooling capability, and submerged valves within containment (Reference 2).

The adequacy of ECCS performance and the staff's evaluation of the applicant's evaluation codel vill be reported in a supplement to this Safety Evaluation Report.

6.3.4 Tests and Insnections i

Toledo Edison Ccrpeny will dcmonstrate. the cperability of the ECCS by d

subjecting all ecmponents to preoperational tests, periodic testing, and in-service testing and inspections. The precperational tests performed fall into three categories. One of these categories consists of sycte:

4 actuation tests to verify the operability of all ECCS valves initiated by Engineered Safety Feature Actuation Signal (ESFAS), the operability of all safeguard pump circuitry doun through 'the pump breaker control circuits and the proper operatien of all valve interlocks.

Another category.is the core ficoding tank tests. The objective of this test is to check the core ficeding system and injection line to verify diat tha Ifncs are fice cf obstructions and tha: the core flooding line d

check valves and isolation valvos operate correctly. The applienne vill perform a lou pressure bicudcun of each corn ficoding tank to confira 1

the line is clear and check the operati.on of the check valves.-

Operational test of all the major pumps comprises the last category of l

testo..These pumps consist of the high prer.sure injection pumps and the

' le4 prcumire ' ' c.(- te ct ree : t.nl pun y.

Oc q%in.;g

4. f t.;

n,..

, a 7%.,

i PD CD D

D l

uv es D "9

. <13.

~

s of these tests to evaluate the hydraulic and mechenical performance of these pe=ps delivering through tha flou paths for c:crgency core cooling.

These pumps will cperate under boch miniflou (through test lines) and full ficw (throu;h the actual piping) conditions.

By =casuring the flov in each pipe, the applicant will make the adjustments necessary to assura that no one branch has an unacceptably low or high resistance. They will also check the systen to assure there is sufficient total line resistance to pravent excessive runout of the pu=p.

Tha applicant must shcu that the minimum acceptabic flows as determined for the FSAR analysis are met by the measured total pump flo. and relat:ivs flev between the branch' lines.

In additien, preoperational flow tests must be conductad to verify the si=ing of the required cavitatin:.

venturies to confirm the as-built ficw split performance of the LPI system.

The system ' jill be accepted only af ter demonstration of proper actuation of all ec=ponents and ef ter demonstratica of flow delivery of all comonants ri

• in d:.c ' p ecui;.-- -.c.

Iba ap;- 11 r.t vill ".:dcrc' rea:J ac peri..'ic te s ti a;;.:f the ECCG cc:'w... :

and all necessary support systems at power. Valves which operate af ter a loss-of-coolant accident are operated through a complete cycle, and pumps are operated individually in this test.

The staff requires that the applicant denonstrate the capability of each notor-operated ECCS v:'Ive to open nd cloca t;.- 8 n, Oc 2 0c:1..n cC.M 21 S c i up.

In " J.'.i:ic -

D rr RW D <l YY %Y 9

3 1

=

__......__....___..__._._...-_...m..

[ in response to a request frca the staff, the applicant has cyclusted his proposed conpliance. tith the positions stated in Regulatory Guide 1.79, "Preoperational Testing of E=ergency Core Cooling Systems for Pressuriced Water Panctors." tiith the exception of the recirculation test under ambient conditions, the applicant has indicated that he vill cenply uith Regulatory Cuide 1.79.

It is car position that a test must bc

~~~ ~ conducted to demonatrate (at c:bient conditions) the capability of the ECCS to operate in th2 recirculation mede.

To avoid reactor coolant syste= contacination, the sttup water may be disch:r cd to externci drains or other systems.

Icnporary arrangements cay bc made to provide adequate sua capccity for punp eparation.

The specific parpose of ', 'is test is to dec.onstrate that conditions (such as inadequate GPSH, ai:-

binding cr vortex formaticn at the sump screens), uhich could adverself af fect ECCS perfomance, do not occur.

6.3.5 Cenelusiens C. :.h bs1: ef cur 2v;'c.r._'cu, c: c ;c.c '. : ;.d

..?

..a _.C

5 for Davis-Bosse 1 plent is acceptabic in retard to a decision con'.crning is: :ance of

.a pecu tinq liwn:;c wit:. th: fclle.. !np, except! aas :

1.

If a small break occurred such t!'at tha high pressure injection (il?I)

I systen alone could replenish the leaking reactor ecolant, the Lcu Pressure Injection (iPI) system uculd be required sote ti::a af ter the accident to provide a water supply fron tha containment sunp.

n.

,re

.. n

^u:

..a

~

~

D D

D<l v a A'

m U _g_

0

.S_

A

_a v

.. _ e

_.;a.~~....~...

s 27-in the crossover lines between the liPI pump sucticn and the LPI peep discharge lires.

Operator actions required to mitigato the conscquence; should be minimi:cd and ECCS component reliability kept at a high icycl. Therefore, the nor= ally closed valves in each of the two crossover lincs should be remote cotor-operated valves with position indication and controls in the control room.

2.

For a break in a core flooding line, a single active component f ailure could degrade available ECCS to the point of compromising the abundant core cooling requirement of General Dasign Criterion 35.

To meet this criterien, Icledo Edison Ccapany has installed a crcos-over line betueen each LPI train which is manually actuated from the control rc.n.

To further r.ini:1:2 cparator actions and..u:.i ize ECCS co ponent reliability, we vill require a passive crossover network between tha two LPI trains.

3.

The adequacy of ECCS performance and the staff's evaluation of the 4

applicant's evaluation ccdel uill be reported in a supplcment to this Safety Eval:ation Paport.

4.

The applicant will be required to demonstrate the capability of the ECCS to c; ; ate in t.:: reci cul: tic-vd 2 c ' ECCS a;.eratisc..

T.

applicant must also damonstrate the operability of the local manual i

~handwhcc1 backup on each ECCS valve pridr to pouer operation.

0 CI07IQ cveuJL m

-q7 o

. S. k..

a a

.....o...-__..

.. ~

15.0 Accident Analvsis 15.1 Gene ral The subnitted safety analysis evaluates the ability of Davis-Ecsce 1 to operate without undue hazard to the health and safety of the public. Two basic groups of events pertinent to safety are investigated by the applicant; abnertal transients and postulated accidents.

15.2 Abnortal Trsnsients The criterien, adopted to assure that th2 reactor coolant pressure boundary integrity is nointained, is that the syste:a pre :sure shall remain belou the ccde pressure li:aits set forth in ASMC Code Se: tion III (1107; :# RCS dati;u precsure). The criterion adopted to et:sure that no fuel dacage has occurred is that the D:'OR must be graater than 1.30 throughoat the trancient.

The applicaat has submitted analyses of abnornal transients and has shoun that the integrity of the RCS boundary Scc been naintaina.1 and that the nini-un 0::5? e:ce.r'ad 1. 30 f o : all analyzed transients.

Tna prest-ura tr.u.si,:.au.u.ich prad ace.1 t...-

' 1@ n t r. : r c.: - r.: c.; N..

.ra -:.trc

- ~

c e n t i f i _.c (bah'-10043) as the ccntrol red withdrawal at low power conditi:ns, resulting in a peak RCS pressure of about 2565 psig. The most severe secondary side pressure transient was the turbine trip fror overpcuer conditions, resulting in a n.a.nimut s teart generator

~

prn so: - c...:b o e t wt.

C::i: -

.;; ch 'n.w 1 ' cf

}.4i

'Pc

.. ntI[i.JJ C-J Tha e;;c 730i'?1. !J r,c. O V *'.1

' ! '. ** ;' ! a n ';

!!C::ll i, I " t.; 2 Ia.

'tOJ G,. ' :

c 'l i f. r c. L n a.

s O

D D

o o E.

~

9 qf l..o((L

) l m

~

.....__-.--..a.w.__w.m....-..._..-._-

The applicant has referenced 3AW-10093 (Reference 3) as their position regarding design features to make tolcrable the consequences of failure to scram during anticipated transients.

We are centinuing our generic review of this area of concern and the staff evaluation of the Sabcock and Wilco:: analyscs vill be sub=itted in a suppleno.nt to this Safety Evaluation Report.

The cer.puter codec "?O*'Z". SAIS" and "PC'P" used icr several ab-r.ortal '. tc.sianto in tha FSAn, ar2 currently under raric. by the staff.

Should modifications to these codes be required, the effect of thcce changes on the Davis-lesse 1 analyses must be censidered.

The cynluation of abnormal transients indicated that the transients presented do not lead to unacceptable consequences and are accept.t'ule for issuanca of the opernting license.

15.3 Accidents The applicant has evaluated a broad spectrum of accidents that might result from postulated f ailures of equipment, or their maloperation.

These highly unliialy accidents, ubich are cescunac.tiva of :c..: spe:. ru:: c 2 :.yp e.; and :hysical leca:1:n; involving the varicus engineered safety feature systers, have bcun analy:el in detail.

The accidents reviewed in the SAR irclude the fellowing:

1 1.

Toss of forced reactor evolant flow resulting from a single reacter coolant pu.mp locked roter.

2.

Main arrauline rutture s

DT Dl 4

V n.

n D

9_

A y

.S.

v

_a

The locked rotor accident was analyzed by postulating an instantaneous seizure of one RC pump rotor. The reactor flow would' decrease

• rapidly-and a reactor trip would oce.ur as a result of a high power-to-ficw signal.

The analysis revealed that at no time during the transient did the DNBR go below 1.0.

The applicant concluded that no severe fuel rod or cladding temperatare excursions are expected to occur as a result of this accident.

The loss-of-secondary-coolant (steam line rupture) analyses has been performed to determine the effects and consequences due to a double-ended steam line rupture. A 36-inch OD steam line rupture, between the steam generator and the main steam isolation valve was analyzed assuning that the reactor was operating at 102% design power prior to the accident.

The present design has only one nain steas isolation valve in each of the steam lines to isolate the unaffected steam generator and to prevent it frc= blowing dry in the event of a steam line rupture. The applicant's evaluation shcus that with a single failure of the isolation valve id the unaffected steam line, turbine stop valves will serve as backup to the first-line isolation safeguard.

The staff requires that the isolation capahtlity of the non-safety grade turbine stop valves be cle'aure-tested periodically and that this test be made a part of

J

_31 the plant Technical Specifications. The staff notes that the worst-e se steam line breaV with regard to reactivity margin was not represented in the FSAR (loss of offsite power not, assumed);

however, the applicant has shown that sufficent safety margin exists

• to justify the differences as n0t ceing significant.

The applicant's evaluation also shows that with single failure of a feedwater stop valve, the closure of the feedwater control vulve and parallel feedwater startup valve will serve as backup to the front-line feedwater isolation safeg'uard. The staff requires that the isolation capability of the non-safety grade feedwater. control valves and startup a.1ves also be periodically closure-tested and that this test be made a part of the planr Technical Specifications. The staff also requires that the feedwater stop valve closure time assumed for the main steam line break (17 seconds) be made a part of the applicant's Technical Specifications and that this time be the basis for all safety analyses requiring feedwater isolation.

It is also noted that consideration of additional single active component failures was not complete. The scope of potential single failures should include tha inadvertant opening of the atmosphere vent valves or turbine bypass system.

It must begconfirmed that these components would provide the largest additional cooldown rate by an examination of all steam line and steam generator active component appurtenances.

After isolation of the main steam line braak, credit was taken for the additional relieving capacity offered by the atmosphere vent valves. Since pressure margin could decrease in the unaffected steam generator if credit were only given for the steam line safety valves (higher setpoint-, the staff requires

e am a.

32-that this event be reanalyzed'with a single failure of the atmosphere valve on the isolated steam generator (failure to open).

A}l the preceding comments on the main steam line break also apply to the feedwater lice break.

The adequacy of these re-analyses will be reported in a supplenent to this Safety Evaluation Report.

References (1) Letter from Lowell E. Roe to A. Schwencer dated Jul'y 21, 1975.

(2) Letter from Lowell E. Roe to Mr. A. Schwencer dated July 9, 1975.

(3) BAR-10099, "B&W Anticipated Transients Without Scram Analysis,"

December 1974.

e

%e