Text

Forwards Revised Pages to Be Included in FSAR Amend Re Containment Peak Temp Profile & Bldg Spray Pump Head Curves
	ML19220C330
Person / Time
Site:	Crane
Issue date:	02/01/1978
From:	Herbein J; Metropolitan Edison Co
To:	Varga S; Office of Nuclear Reactor Regulation
References
	GQL 181, GQL-0181, NUDOCS 7904300415
	Download: ML19220C330 (23)
	v • d • e

.

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gz n,- n -wm METROPOLI FAN EDISON COMPANY POST OFFICE BOX 542 REACING, PENNSYLVANt A 19603 TELEPHONE 215 - 929-2601 rebrua y 1,,, S

-ji GQ,1 0131

,/\\'y"k.ml@A Diractor of :Tuclear Reactor Regulaticn

,.h' Attn:

Mr. Steven A. 'larga, Chief M7 p -. Luy[g gg

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d Lj;ht '4ater Reacters 3 ranch :ic. k 3o i;. S. :!uclear Regulatory Cennissicn

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Washingten, D. C.

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My h/ [UM

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Three hile Island :!uclear Station Unit 2 (CG-2)

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C.-P,-t o Dccket :Ic. 50-320 Containment Peak Temperature Profile and 3uilding Spray Pu=p Head Curves Tvc letters (GCJ 0139 and Olk0) forwarded to you on January 24, 1978, enclosed a:_ ended and new pages of the T:C-2 FSAR ccncerning the Centainment Peak Te=-

pera*are Profile and Building Spray Nnp Eead Curves.

Several of these pages required furtha-a-anding and are to replace these previously forwarded in the above referenced letters.

Included in the attachments are other pages that vill also be incorporated into a future amendment. ~hese are being provided to help expedite you review.

If you have any questicns, please contact ne.

"'ncerel,

I-G. Herbein v.

Vice h esident-Generation u G.,,,. : u r.o : c, g n

a Attachrents:

TC-2 FSAR pages 6.2-23, 23a, 23b, 68, 153-15, S34 2-21, 22, 7.3-1, 3, 5, 8.1-9, S.2-2, S.3-16, 15, IT.lan, S3-22-23; Fir 2res 153-10, Sheet Ta of 8 through Tf cf 3.

c:

Mr. Harley Silver "ivision of Reactcr ~.icensing U. S. :Iuclear Rer21 story Cennissien n-o% j

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Washington, D.C.

20555 u -

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7901300'/6 1/ /

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ill wetted materials in recirculation flow path arep ttible with reactor coolant.

The ajor components are constructed d. 'swainless steel

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with 'the exception of the sodium hydroxide storage tank which is constructed of carbon steel.

Some carts of major components, such as pump seals and valve seats are constructed of other corrosion and radiation resistan: caterials.

6.2.2.2.1.1 Redundancy and Independence of Components and Systems Cooling of the resctor building atmosphere is assured by either the reactor building spray system or the reactor building air ecoling units.

Adequate cooling of the reactor building af ter a LOCA can be accomplished by (1) full capaci y of the reactor building spray systen, (2) full capacity of all

he reac:c: building air cooling units or (3) tuo of the five reactor building air cooling units plus one-half of reactor building spray system.

Redundancy of the reac:c building spray and RB air ctoling units, and redundancy within each systen, assures there will be no loss of the R.3. cocling function.

Sodium hydroxide chemical addition is assured based upon a dual spray syste chere one of the two R3 spray subsystems will provide the required engineered safety features system function.

6.2.2.2.1.5 Design of Recir:ulatica Piping Dual recirculation lines connect the R3 surp via isolation valves, with the suctice of the Reac:or Building Spray Pumps.

Each recirculation line is run inside and concentric with, an outer guard pipe thus providing a double barrier against possible outleakage.

The guard pipe at the sump end is welded :o the outside of the sump liner.

At the outboard end, the guard pipe is at: ached to a jacket which encieses the sump isolation valve to the outer surface of the pipe on the downstream side of the valve.

in this canner, a double barrier encloses the recirculation pipe and its isolation valve.

Pro;isions are made for leak testing of the space between the double barriers.

6.2.2.2.1.6 Net Pcsitive Suction Head Assessmen:

The N?SH available f or the Reactor Suilding Spray Funps has been calculated for both design and runcut flou ccnditions.

The NFSH required is based on vendor perf ormance test data.

The tinimum required N?SH for the Reactor Eu11 ding spray purps has been determined based on the runcut flou (1740 gp=) assuming that operator action has been taken prior to switchover to the recircilation made.

The laces: available NFSH fo: these purps occurs when they take suction f rom the Reactor Zuilding su p.

Based on cal-culations in accordance with the requirements of Regulatory Guide 1.1, the available N?SH for this condition is 18.6 feet.

- H. - H m NPSH = ?A+EZ r

\\t For the limiting situatica, when the pu=ps take suction f rom the Reactor Building surp water at a tenperature of 25CF, the values in the equation are:

3. ;. p.y. p =. _ _.,_4_.__,,.__.. o. o -.._.,

oe n-r s<

l l' Fiping friction losses E

=

7 E, = Elevation dif ference H,.n.= Fluid vapor pressure cased on surp terrerstare

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o._}

Initially, the syste Operates taking suction free the 3erated h*ater Storage Tank.

During draudown and accident conditicas, the Decay Heat Renoval pu=ps, Reactor Building Spray purps, and Makeup pumps are supplied frem the 3 orated Water Storage Tank. The li=iting case NPSH requirements are calculated assuring that the Decay Heat Receval pu=p on the same power / safety train as the LOCA break runs out to 4150 gpt, and the other Decay Hea: Receval pump feeds the intact line 3000 sp rated flew. The Reacter Suilding Spray pumps supply 1500 gpc each, and the Makeup purps runout to 550 g;= each.

These assu=ptiens taximize NPSH-Af:er no less than 35 cinutes of cperatica, the control Roc: receives a low-lev level alar free the Borated Water Storage Tank, and the syste is switched auteca:ically to recirculation, with the Decay hea: Re cval and Reactor Suilding Spray purps taking suction frc the Reactor Building surp and recirculating the sprayed water. The two lines frc the surp feed one Reac:cr 3uilding Spray purp and one Decay Heat Recoval pump, respectively.

Table 6.2-11 provides the result of the N?SH assessrent of the Build-ing Spray (35) and Decay Heat Receval (LP1) pu ps.

The design cal-culaticas supporting the result s hava been verified by:

A.

Measurecent during plant testing - ficw res:s of the LPI dis-charge piping and 3S/L21 cceron suction piping wara perforced at reduced fleu.

Test results shcued that calculations over pre-dicted pressure d: cps by 9% for one of five data points and less than 4% for the retaining points. The syste resistance was developed for flou frcr the bora:cd water storage tank to the

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The resulting flew con-diciens are 1700 gp for 35-?-1A and 1740 gp for 3S-?-13.

Based on these results, syster resistance and systen flow calculations were assigned a plus of cinus 10% =argin as conservatively appropriace, givfng a canimum flew of 1760 gpc for 3S-?-1A and 1800 sp for SS-?-13.

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Scale ecdel testing - surp codel tests were perforced to determine if a tendency for vertex for ation existed in the Reactor Euilding surp (see Section 6.2.2.2.1.9).

These' tests aisc provided cc parisons of experizental vs. design pressure drop calculatiens.

NFSH calculaticas were adjusted as a result of :his ce=parisen to ore accura:ely acccunt for losses thrcugh the su=p screening and portiens of piping that eculd not be field tested. The entrance loss coef..cient for the suction lines has been taken as 1.25 based on extrapolatica of the ecdel test results for the suction line exhibiting the higher entrance loss.

The screen less coefficient was increased an additional 15% to account fer screen blockage effects.

This reduction is censistent with codel test results.

The pressure drop in the approxi=atelly 60 foot secticn of 18" piping between the end of the su=p and the 3WST has been calculated using values of loss coefficient for piping and gate valves (DE-V6A/3) in this line.

These values are conservative as discussed in A(a'ocve).

Since testing is not possible with water at accident temperatures, available NFSH dur'ng accident conditions has been verified by cc=-

putatica based on the pressure dreps which were either seasured during cold testing, cbserved during =cdel testing, or cceputed for sectiens of piping not included by either test cethed.

As indicated in Table 6.2-11 the L?I syste: pumps have an adequate NPSH =argin when drawing sue:ica free either the SUS A cr surp.

ne building spray pumps exhibit adequate cargin when drawing wa:er frc: the EWST.

As described in Section 5.3.1.17, valves 3S-VIA/3 will be throttled ::

limit flew, assuring adequate NPSH available to the pu=ps in the re:irculation code.

Figures 6.2-23 A/3 presen:s the results of the building spray syste: N?SE assessment and includcs systen resistance curves which were developed facecring in all of the catheds discussed above.

The building spray pu ps testing vere :ested to confirm the pump N?SE required using the Hydraulics Institute Standard for pt p testing.

Fu=p operation in a li ited suction ecde has also been conducted during NPSH testing.

Although no extended time test was perforced, the purps were cperated for approni=2tely 15 minutes at several ficw rates with head reductiens of up to 9% with no undue vibration.

De-flection readings during operation with 9% head reducticn were apprcxicately 1.5 tils.

Further= ore, the pt:p manufacturer has verified that the purps could operate at hend losses between 5% and 7% for 3 to 5 days.

Piping leading fro = the sung to and including DH-V-6A/3, was rain-tained clecn throughcut field erection under QC surveillance and will not be unsealed withou QC surveillance until construction in the area has been ccepleted.

In additica, the sump lines and DH-V-6A/3 S }

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Table 6.2-11 NPSH ?.9 i

.quIREMENTS FOR THE REACTOR SUILD...G SPRAY ?t"PS AND THE DECAY HEAT REMOVAL PUMPS Decay Hea: Renoval Pu=ps Reactor Building Spray Pumps Suw 4.c.

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Assumes temporary strainer in suctic-line to pumps has been removed.

c.

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}'a::inur calculated DH flat. cecurs.i:h a pos ula:ed break in the core floodine,.

line be:veen check valve CF-V3A am. the :::c flood no::le, mith ficw con::ct valve (DH-V-12SA) fully open.

Long :er cooling is provided by opening DE '. -7 A o r 3 to ciiver t a fraction of the discharge of the Decay Heat Re cyal pump :: the sue:ier of the S'akeup i.'a p urp s.

The sectica requirements of the E pumps do not change signi:ican:1:., am. a suctica head of 43.2 f ee: is v

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Containment pressure analyses vere perforced with the C0"TE'2T cede using the folleving assurp:1cns.

Condensing heat transfer was

=edeled using the "AEC-Tagari Correlation" described in the topical report 3AW-10095, Appendix A, Part 3.

The Uchida & Taga 1 correla-tiens were used vi h cultipliers F-and F, (described in above reference) set equal to 1.0, and, tee ini:Isl hea: transfer coeffi-cien: F3 rese: to 5.0 ETU/hr-f: -F.

Fcr all cases, the temperature used in calculating condensing heat transfer Oc passive heat sinks was the containment atmosphere terperature, which can be super-heated.

The analytical =cdel used to account fer redistribution of

= ass batveen the liquid and at csphere regions in CONTE::?T is descr. bed in See: ion 3.9 of EAW-10095.

The Babecck & Wilecx ver-sicn of the CONTE"?T code. as described in the Topical Reper: SAU-10093 does not contain a mass / energy =cdel for localiced condensa-tien on passive heat sinks er fan ecclers. A :sss/ energy balance in the containment at:csphere is used to predict bulk cendensation when desuperhea:ing eccurs.

In the analyses perforced, all conden-sate was assured to go directly to che surp region.

Condensing heat transfer to passive heat sinks is based en experi: ental data (see 3AW-10095).

In order to deter ine the effec:s of elevated containment te=pera-ture, the utrst case accident (:urbine s:cp valve failure) was analyced in tre ph:ses.

First, the worst case temperature profile was determined.

Specification cf the condensing stean Uchida correlatica was fcund to provide :he ecs: severe profile.

Hea:

transf er into the equipmen: was caxiciced by choosing the ecs:

conservative of forced convection hea: transfer, four times Tag 1:1 or four times Uchids.

End cf bicudern was specified at 50 seconds eith respect to using the Tagari correlation.

Forced convection was calculated using the Frend 1 correla:icn:

u= 0. 0 7 0. c' e,e. 8. 0. 3 3 0

3 with seca velecities of 30 ft/sec chosen in calculating the Reynolds number for the pressure sensor enclosures, motor boxes, and containment electrical penetrations.

A stea velocity of 80 ft/sec.

was used for the contain:ent fan coolcrs.

This velocity corresponds to their nor:21 accident flev rate.

Since the equipmen can caly be ecdeled in one di=ensica, the test critical geccetry or =aterial was chosen. For the instrument enciesuras, electrical cabling, and cc:cr boxes, ne choice vas required since these are bcited retal cylinders and bcxes of :niferr thickness.

The cen-tainment fin eccler is a metal c3 1inier eith fins.

The fine h2ve been 1su od m

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c.

Cont ainmen t isolatica valves, plus the decay heat removal valves Selected electric power and instrucentation cabling and ter inal connections e.

Pressuri:er safety valves f.

Trans it:crs for the follouing instrucentatien:

1.

RCS vide range pressure

11. Stea: generator pressure lii. Pressurizer level iv. RCS het leg and cold les temperature v.

Stea: generator level.

Qualification tests have been perforced cc each of the above ce=penents as described in Sze:icn 3.11.1.1 and 3.11.1.2 of the FSAR and EA','-lCCO3.

Addi:icnal inferna: ion concerning the containment electrical penetra: ions has also been supplied to the Staff.

Althcugh equipment was, in general, qualified for a long ter:

2So?, ICOT: humidity envirennant, sc:a cceponents were exposed to higher te:peratures in the course of their envirennental qualifica:ica. The containten: elec:rical penetrations vere qualified in three sequential phases. The first phase exposed the penetrations to a 3407,1C3 psig LCCA environ =ent for four hcurs.

Valves inside centainten: ucre tested for one hour at 329F at 90 psig.

These cotors are enciesed in NEdA IV boxes, so that the te perature transient seen during a stea= line break is a surface phenacenen en the box, and is act actually experienced by the =ctor.

The only contain:ent isolation valves inside containcent which are not noter operated are the containuen:

purge valves, which are hydraulically operated. These valves receive a 4 psig containten: isolatica signal and will be closed before the centaincent vapor temperatura exceeds their qualifica:icn temperature.

Cabling has been qualified.in a LOCA environ = ant at 350F for 10 hcurs.

In addi:1on, cne cable venicr has certifie,i tha', gi:

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1 7.

3.1 DESCRIPTION

The interaction of the SFAS and the.envineered safe:v features systen is shown o

On Figure 7.3-1 consisting of; a.

Sensors used to de:cc: abnormal conditions.

b.

Ac:uation logic that takes signals frem the sensors, a.

_nges then in two cut of three logic, and when abnormal conditicas exis,, provides signals to those actuation systens used to mitigate the effects of the abnormal conditicas.

c.

Actuation systens tha: mitigate the effects of design basis events by isolatica, heat removal etc.

d.

These support systems tha: must be present for the correct operation of the actuation systers.

7.3.1.1 Svsten Descri:tions (Safety Fe2:ures Actuatica and Encineered Safetv Features Svs:ery) 7.3.1.1.1 Safety Features Actuation Sys:en tne safe: f eatures ac tuation system (2:..a > 1s ts.e instrucentation man 4. coring and trip initiating systen for the ESF, i.e.,

i: does not include the final equipmen: elements, such as pumps, notars, etc.

All portions of the SFAS are required for safety.

.a,.,.,..,.2

,n;._,a 4. ng v...ui.-

n:

.m

.s The SFAS renitors variables :: de:ect irss :f in:egrity in the reactor ecolant sys:er pressure boundary.

Upor de:ection of ou:-of-licit conditions of these variables, it initiatas opera:ica of the high and let pressure injection sys-

ens, :he reactor building iscla:ica and ccling syster and the reactor buildin; spray syster as summariced in Table 7.3-1.

In addition, it starts the emergency fiesel genera: ors chich are leaded if 21; 230 r. sources of power are lcst.

The devices actuated by the SFA5 are listed in Table 7.3-2.

The SFAS itself extends frca the sensing instruments to :he final actuating devices, such as circui: breakers and motor contac:crs.

The reactor coolant pressure and reac::: building pressure are 5e variables which initiate safety features action.

A 10 pressure of 1600 psig in the reactor ccolant systen and pressures of s and 2S psig in the reactor building l

are the levels at which the various types of safe:y features actuation are initiated.

These are hereafter referred to as levels of protection.

7.3.1.1.1.2 SFAS Logic Ac:icas initiated by the SFAS are so- : riced in Table 7.3-1.

Two-aut-cf-three logic is used as the basis far ac:catica.

The cutput signal of each reactor coolant ressure sensor and reacto-b'4' ding,ressure sensor is O

V Udb

,_i-

The continued bypass of a channel by opera:Or action is possible only after a trip of one of the two remaining pressure switches.

Deenergicing the output relays of two out of the three channels initiates reactor building ecoling and isolation and opens all valves required for reactor building spray.

The valves in :he reactor building spray'syster are opened by a two-out-of-three matrix formed by a set of pressure switches sensing a building pressure higher than 4 psig.

One reac:ce building spray pump is initiated by a two-out-of-three = atrix forced by a set of three pressure switches sensing a rise in reactor building pressure to 25 psig.

Another set of three pressure switches l

1s wired similarly into the star:ing circuit of the other spray pump.

The SFAS is basically a three-channel redundant systen employing two-cut-of-three coincidence between ceasured variables.

The failure of one of th

" ree channals will cause that channel to trip leaving a one-out-of-two logic wen-figuration.

Thus, the ability of the SFAS to perform 1:s assigned function will not be impaired.

cv The reliability of the two-out-of-three logic is =aintained by providing a separate two-ou:-of-three matri:, f or each saf e:y f eatures auxiliary.

The loss of vital bus p:wer in the instrument strings will initiate a trip of that portion of the logic associated with the affected instrument string.

The loss of any one vital bus which powers the systen logic will not initiate syster actuation.

The loss of two ou: of the three buses used, however, uill actuate all safety features excep: the building sprays.

Eeroving modules from an instrunen: stri.:g till initiate a trip in tha: portion of the logic associated eith the affe::ed instrumen: string.

Removing logic modules from one nrctec:ive channel does not prevent ans other prc:a::ive chanac1 fren initiating sys:er action.

Provision has been made for :est and bypass of each trip bistable independen:17 as described in 7.3.1.1.1.3 and 7.3.2.5.

7.3.1.1.1.3 Bypasses The trip functions of the energenc, core injection actuation signals can be bypassed whenever the reactor is to be depressurized belot the trip set point of the bypass bistables.

Sypassing cust be initiated manually for each channel within a fixed pressure band above the protective systen bistable trip point.

The reactor coolant lou pressure safety feacures actuation signal may be bypassed only when the rec ter pressure is 1,820 psig or less and control pcuer is available.

The bypass is au:: a:ically removed when reactor pressure exceeds the 1,820 psig value.

A anual bvpass reset suitch is also provided for return to normal operation.

In additier, manually actuated switches are available for tripping each charnel.

"7 O *{

t

)G JUJ 7.1-

9 There is redundan:) in ESF equipment such as cotors, pumps and valves as dis-cussed further in Chap:cr 6.0.

7.7.7.1.1.7 7'.v.,.b.'.'

a Dive rs i:) is provided in pro:ec:ive a :icas and tetheds.

For example, there are two separate se:s cf si;nals f:r ini:ia:ica of the energency injection sys-tems, one fr : reac:Or ccolan pressure and cac frc reactor bui? ding pressure, and these paraceters are ceasured with differen: types of instruments, the coolan; pressure wi:5 ::aascitters and bistables and the building pressure

.4.,

s..

..;es.

. -.., e s A.,.

Also, manual rip switches, which are independent of the auto trip instrumen-tati:n, previde for further diversity.

A manual trip pushbutton has been pr:vided on the con:rol roc: =casole for each of the 1600 psig and 4 psig 3 eve _s c:. protect :r a:.

eacn a::uation.

Operation of the pushx.utten energizes relays *: hose con:ac:s perfor an "0R" function with the matrices of the auto-matic actuation.

The p ver supply for the manual trip relays is takeo fre he s:ation batteries.

?cwer used for the :wo act.uations is provided fro =

s e,e,. 2._,., a t.. < a o. os

.24

,assm.a o..,..,.

s 4

e l

The 23 psig level of protectica is manually testable frer the SFAS conitoring l

panel in the con:rcl ::: by actuating One pus'. button to trip a channel, or two pushbu:: ens to star: a spray pump (with spray ralves closed and test valve c

e.. =. ; ).

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e 3

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o building pressure si;nal.

If the rea :or builcin; pressure increases to 23 psig l

the spray pumps vill be star:ed tc reduce the pressurc cad temperature of the

.c, c.,

u. 4.1 2 4.3 a _,s,r 3.m.

w...

The reactor buildin~ sc. ray purcs are actua:ed b.

two sets et threc pressure o

suit acs each-The swi::hes are calibra:2d to close a: 23 psig reac:Or building l

pressure.

Each set of stitches is associated :o one pump through a two-cut-of-three 1cgic.

No by-ousses have been provided to the logic in this sys:cr-

,.3.1.1.3 Reac:ce 3cildin; Air Ccolang Syster Th : reactor bu lding air coclin; sys ter is sh' cr

ore 6.2-27.

The logic 50.

his sys!er is 7CT: o f t h; ' psig reaJ:.'r build. w prCss"re trip and 'h2 s;1.S...

s des-.4.

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g._9 7

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TASLE S.1-2 (CONTINUED)

v. 0. _: 2.

1.

The above represents the heavies load on Diesel Generator DF-X-1A.

Loading for Diesel Generator DF-X-13 will be the same as for DF-X-1A except that some A 5 3 loads are interlocked in 1 out of 2 logic with "A" type loads preferred.

Thus, generator DF-X-13 will be approximately 50 Kh' less loaded.

2.

Denotes loads based on calculated break horsepower require-cents.

Other leads are based on motor name plate rating.

3.

- Deno:es loads other than no::r loads.

4.

These values do not include heat tracing and main steam valves MS-V43 and 73.

5.

NR-V2A is energized by S:R-?-la af:er time delay of 10 seconds.

6.

If MU-?-1A fails to start, MU-F-13 nili start at time 14 seconds.

7.

The emergency feedwa:er pumps are au:cca:ically started.

Although not needed during a LOCA, no credi: is :aken for turning the off until 30 minutes elapse.

3.

The choice of 40 minu:es for sh:uing initial loads and subsequent loads is based on the fact that 'O ninu:es are required to empt'; the borated cater s:orage tank after a double-ended break in the largest pipe in the reac:or coolant syster.

This tire is lenger for smaller breaks, and can be as long as 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

9.

Equipment listed as "Manua." is azailabic fer operator or automatic actuation after 100 seconds.

10.

1: AH-C-16A fails to start, AH-C-163 will start at time 20 seconds.

l 11.

If AH-C-17A, AH-C-19A and AH-C-20A fail to start, AH-C-173, AH-C-193 and AH-C-203 will start at tire 28 seconds.

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3.3.1.

.2 General Cable Considerations In general, notor and transformer feeder cables are rated on a continuous a.

basis a: 125 percent of full load current.

This provides for occasional motor and equipment operatica at service factor ratings.

Power cables installed in cable trays are specified to be one layer deep only, b.

Fire barriers are used at cable :rr s and cable runs where they enter or leave an IEEE ::o. 30S Class IE area and vhere vertical trays pass through floor openings.

A-C power circui:s within the plant are prote 'ed by circuit breakers.

c.

D-C circuits are protec:ed by fuses and/or circuit breakers.

d.

? ver and con:rol cable trays over 6 in. in vidth are ladder type.

Where there are horizontal trays passing under coe, gratings, :Se top tray has l

a solid cover.

All exposed sides of vertic.

trays passing betueen floors have solid c: vers to 3 ft. above their ficer jenetratices.

Power cable a paci:1es are de ermined en the basis of the maximum ambient e.

tempera:ure expected, the ele::rical current requirenents of the respective equipment and the shor: ci.cui: current protec:ica requirements.

f.

An ambient tempera:ure of 50C is assumed for :he in:eriors of the Reac:or Building, the Diesel Generator Building, elevation 2i8'-6" of :he Auxiliary 3uilding, and the area above elevati:n 317'-0" of the Fu e _' Handling Building.

All c:her are2s are assured to have an anbient ter; era:ure of 400.

These arbient te pera:ures are the design basis for :'l power cable ra:ings.

s

.,7

-)

/

V n0.7_..

s

9 1han needed, instrumentation cables are shielded to minimize induced voltage and magnetic interference.

L.' ire and cables related to engineerec safety fea:ure 2nd reactor protective systems are rou:ed and installed to maintain :he integr2:3 of their respective redundant channels and pro:ec: them from physical damage and building sprays.

Liquid tight flexible condui:s are used to terminate conduits at

-c..s-4.,.. w.e. 3 o m,. 4. -...

b.

Cabling for radundant ec:penen:s has been identified b, four basic different colors and the redundant pcuer, instrumentation and control cables are run separately.

1.

There is four channel separa: ion for the reactor prctection and three channel separation foe the saf=ty featurcs actuatica circuits.

These separations are maintained from the sensor :hrou5-h the analog racks to the logic or relay cabinets.

2.

Control and pcuer cables for the operation of redundant components 3

in sar.ety reatures systems are rcu:ea separate y.

c.

Separation is maintained for safety features redundSn bus d-c control a.e,_a. s.

-,.u,.

.n.

rs e,-3...e e.ed s o,.: e t.

d.

r.,.,,., b u,.<., a t,,,

ne

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.32 feature equipmen are routed ;y separa:e paths.

Miniuu: separation be-u g. - t. s, s. <. -..;. g :o..c.au.>.a..e.

,, g 4.,. e c,. g. s, :,.. ;

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.c gated to ensure the single fai ure :ri:eria is not violated.

Trays containing cul:i-conductor pouer cables are one layer deep.

Trays are bonded by bare copper to the grounding 1.op to provide a lou impedance return path for ground currents.

Cables of different voltage classes, uhen running together in a tray, are separated by natal barriers.

Trays contain'n; control or ins:rumentation cabling are filled to not exceed the appearance of 100; fi.11.

Tr-"s are bonded by bare copper lov ingedance reture path for ground to the grcund loop to p r a '.' '. e m

currents.

Cables running in cable :r2.cs arc rcuted :brou;h specific trays by a

d

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printed pul.1 slips.

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3

.31

(3)

"PCD Internal *.udits The Sr.s'-S?GD Reac:O r, Ne.? Products and fecvices, Purchasing and Quality Assurance Activities are subjc:: :: periodi: internal audits

o determine :he extent of implementa:1:n of SPGD procedures as they affect the quality displayed by these er;2nizati:ns in executing com-nercial nuclear steam sys:er, componen:, rnd fuel contracts.

Subse-quent audits determine the effectiveness of management approved cor-rective a: ica :o prior discrepan: findin;s.

These ac:ivities are audi:ed to :he foll uta; elements, as applicable:

design centrol; procure:en: documen: centrol; instructions, proce-dures, and drawings; document con:rol; con:rol of purchased asterial, equip =ent, and services; :orrective a::icn; and QA records.

17.1.13.3 35R A comprehensive systen of planned and de:unen:ad audi:s of the ho c-office and site Engineering and Quality Assurance activity are carried out to verify ccepliance with all a :1vities affec:ing quali:y.

Audi:s are peric :ed in accordance with written procedures and checklists by appr:pria:ely trained personnal not having dira.

responsibility for the areas that are audited.

The Director of Quality Assurance is rer,.cnsible for *he timely initiation and e

c. a.is. ou.24.s.

a.4.,-

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.a u.

lis:

hi:h covers as a minitu :

base.ine docuren:3, design criteria, calcu-Ic:icns, bidder qualifica:icns, dee -i r drauings, vendor drauings and data, change centrols, FSAR deviations, design :Inss.:ications, contractor quality con:: 1 evaluations, project records, and Engineeriag and Quality Assurance prc:edures.

This group performs 2 df's to ensure : hat design control pro-cedures, specifi:ation c ntrol p: =clures, and 0:'.ier procedures relating to engineering control are being innlamsnted.

The audi: checklis: dealing wi:h the Engincering activi:y, accc=panied by the audi:

can's report, is submit:ed to the Dire::c: of Quality Assurance with copies to the Project Manager and Vice President, ? cjet: Operations.

The checklis:

dealing with Quality Assurance acti-tity is submi:ted to the Director of Quality Assurance with a copy to the Quality Assurance Group Supervisor.

The Project Manager is responsible for reparting the corrective actions to the Director of Quality Assurance. 'ti:5 a ecpy :: the Vice President, Project Operations.

The Quality Assurance Group Superviser is responsible for reporting the correctice actions taken to the Direc:or of Quality Assurance.

This report besides including the ;crre::ive action taken also indicates the responsible parties and the time limits for impl nentation.

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17.1

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tested during normal pcuer operation of the rea :or.

Also, for each piece of equip en: c: conponen: identified, specif, and justify the

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reactor; however, a non:hly : n;;ional :es for the ESFAS Ac:uacion Logic and Manual Ini:ia: ion is required.

Due :o the design of the

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Cecay Heat Valve, DE-V2.

See FSA?. See: ion 9.2.2.1.4.

2.

huelear Services Closed Cealing *iater ?" p ':5-?-10.

See FSA?.

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u.,. =. e,r t., 3

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ns contacts cannot be ve.ified unicss Diesel Genara:or Protecti.ve C.4.. u.' **,<

4s..".."'.'.,

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,..' n ua.1.' )-

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..f operating the associated lockouts.

3.

DF-X-1A and DF-X-13, Volta;e Shutd;vn, Prc:ective Relaying 31ocking conta::s.

These con: acts canne: be verified unless D i e s e1 ',s'e.e a *. o. D. o.e t. ' va C.'.-"'..,"#-

..... " '. '. ' ", *. d.

ee A using junpers or by nanu;11y oparating the as ;ociated lockouts.

6.

CF-X-1A and DF-X-13, redundant startins cor acts, both Diesel Gene ra tor :.) starting con:ac:s rc in :nc sana dt.S test group.

Therafore, it can be veri:12d tha: the Diesel Cencra:nr star:s; howcVer, I: cacao: be verified whi;h conta:: star:cd the diesel.

E uicsel Generator Auto Loadin;; Sc.;ucr :.a ; Safcty.~e:.

,cas Ac:uation Sys:cn inizia: ion Con:: cts.

Seu 75.;A :cc: ion 3.3.L.1.S.6.

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>>Jn: ir critical.

Sec Section 6.3.3.

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3.1 DESCRIPTION

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