ML19206A360
| ML19206A360 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 01/19/1978 |
| From: | Herbein J Metropolitan Edison Co |
| To: | Silver H Office of Nuclear Reactor Regulation |
| References | |
| GQL-0076, GQL-76, NUDOCS 7904190385 | |
| Download: ML19206A360 (21) | |
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rer* x.n **ocons METROPOLI FAN EDISON COMPANY
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PCST OFFICE BOX 542 AEACING. PENNSvtV ANI A 19603 TE: EPwCNE 215 - 929-2601 January 19, 1973
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d.c) d Director cf :iuclear Reacter Regulation
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Attn:
Mr. Earley Silver Divisicn of Reacter licensing
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U. S. :Tuclear Regulate y Oc==issica L
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s Tnree Mile Island :!uclear Station Unit 2 (2C-2) license 'Ic. CFPR-66 Oceket :To. 50-320 FSAR Arend ent :To. 62 'nfor._aticn In order to expedite your reviev, enclcsed are advanced copies of pazes which we anticipate vill be included with submissicn of Amendment IIc. 62 to the CE-2 FSAR.
Should you have any questicns, please contact me.
Sincerely,
- 3. Herbein
' lice President-Generati:n w b.:. U.:m. u 6
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Faces 1.1-1 6.3-30 1.1-2 7.5-6a 3 11-3 8.2-2 3.11-3a 3.2-2a 3 ". ~ )-
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..s...v L v o..v This Final Safety Analysis Report is submitted. in support of the applicaticn of Metropclitan Edison Ccnpany (Met-Ed), Jersey Central ?cver & Light Cc=pany (JCP&L), and Pennsylvania Electric Cc=pany (Penelee) for a Class 1 Chb license to operate the nuclear electric generating statica designated as Three Mile Island Nucla - :'atien Unit 2.
Subject to the approval of State and Federal a
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to Unit 1 en '"..ree Mile 'sland nea-the East shcre of the Susquehanna F.iver ic.
Londonderry "ev. ship, Dauphin Count /, Pennsylvania.
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5 d's operating er under construction.
It uses chemical shis and centrol rods
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ry Babcock & Wilecx is also supplying the fuel fer the first core of the reactor.
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3.11.2.1.3 Reactor Building Air Cooling Units A prototype test of the fan =otor asse=blies of the Reactor Building Air Cooling Units was perfor=ed in accordance with the Air Moving and Conditioning Association Bulletin No. 210-67.
The prototype test was perforced under environ = ental conditicas equivalent to the post-LCCA Reactor Building Environ =ent.
A discussion of the test and results is contained in 6.2.2.3.2.2.
3.11.2.1.4 Power Cables Power and control cables are designed :o operate properly in a te=perature range of -5F to 100F and to withstand, without inj ury or dacage, an integrated dose of 5.6 x 10/R.
The vendor har supplied guarantees that cables will operate properly under these conditions.
In addition samples of the power and control cables have been tested in simulated post-LCCA environments to de=enstrate suitability for service.
Ihese tests involved exposing the cables d
to an integrated radiation exposure of 1 x 10 R, sprays of dilute boric acid, sodiu= hydroxide and sodiu= thicuulfate, and te=peratures and pressures exceeding DBA conditions.
3.11.2.1.5 Penetratica Asse=blies Electrical Penetrations have been designed, tested, and docu=ented in accord-ance rich IEEE 317 (1971 for low voltage and 1972 for =edium voltage), IEEE Standard for Electrical Penetration Asse=blies in Contain=ent Structure for Nuclear Power Generating Stations. The penetrations were designed to operate continuously in a nor=al operating environ =ent of 110F, at=ospheric pressure and 20 to 100% humidity, and in =axi=um e=ergency environment conditions of 286F, 60 psig and 100% hu=idity for two hours.
The assemblies are also designed 7
to withstand an integrated radiation exposure of 2 x 10 R without da= age.
In addition, transition field splices will be provided on all power, instru=ent, and control circui:s penetrating contain=ent which will be required to operate during post-accident conditions. The field splice insulating =aterial has been subjected to a test program to si=ulate normal service, a loss of coolant accident (LOCA), and cooldown post LCCA, and which included si=ultaneous stea=, che=ical spray, and radiation exposure.
These splices will be enclosed in junction boxes which have been designed for NEMA 4 class service.
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02- 02 (1-2L-T3) 3.11-3
3.11.2.2 Equipnent 'a'ithin the Auxiliary and Fuel Handling Buildings Since no abnormally severe environ = ental conditions are expected in the auxiliary building during and subsequent to any of the design basis events, the normal and customary industry standards and tests as discussed in Chapter 3 are considered adequate to assure acceptable performance of safety-related equipment.
3.11.2.2.1 Instrument Cables The design and qualification of instrument cables is the sa:e as described in 3.11.2.1.1.
In instances when the predicted environ =ent of cables for specific circuits does not require the characteristics described in 3.11.2.1.1, substitute cable which will ceet the requirements of the predicted environ =ene will be specified.
3.11.2.2.2 Power and Control Cables The design and qualification of power and control cables is the same as described in 3.11.2.1.4 3.11.2.2.3 Motors All motors have class B insulation. Motors are designed to operate in an ambient te=perature of 40C.
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.. \\ > is 3.11-3a A2. 52 (1-2L-73) l
6.1.1.2 Contain=ent Isolation Valves Isolatica valves provide a double barrier so that no single, credible failure or =alfunction of an active co=ponent can result in leakage due to loss of ccatain=ent isolation.
The installed double barriers take the form of various types of isolation valves located on each side of the reactor building penetration in piping syste=s which have portions both inside and outside the reactor building.
All isolation valves inside or outside conts' en: which are not nor= ally locked closed and all re=otely operated isolation valves inside the con-tainment have position indicators in the control room.
6.1.1.3 Heat Re=cval Syste=s 6.1.1.3.1 The Reactor Building Spray Syste:
This system serves as engineered safety features function, along with the reactor building air coeling system, to ecol the reactor building at=osphere following a LOCA, reducing pressure within the building, and minimicing the potential for leakage of radioactivity fro the building to the environ =ent.
The spray syste= also removes radiciodine and suspended radioactive partic-ulates. The fluid for the spray systen is supplied frca the barated water and sodium hydrcxide storage tanks and is discharged into the reactor building by a pu=p in each of the two spray system circui:s.
The system consists of two circuits each containing a 50% capacity pu=p for heat removal frem the reactor building atmosphere.
Each of the two separate reactor building spray pu=p suction headers is connec::d to the borated water and sodium hydroxide storage tanks, and to the reactor building su=p.
Each spray pu=p will take suction initially from the tanks and subsequently from the su=p.
6.1.1.3.2 Reactor Building Air Cooling System The Reactor Building Air Cooling Syste= consists of five (5) units connected to a cc==cn system of ductwork for air distribution.
Each cooling unit consists of finned, water type, cooling coils and an electric =otor driven axial flow fan.
The five (5) units are asse= bled in a cct=cn =etal housing with suitable divider places and back draf t dampers to permit shutdown of selected fans without affecting the air flow from the others.
For nor=al operatica, cooling water is supplied to the cooling coils frc= the evaporative coolers with three of the five air cooling unit fans in operation.
For LCCA conditions, four fans are operated, and the cooling water supply is drawn from the nuclear services river water system.
This LCCA cooling water ficw is provided by two of the four reactor building e=ergency cooling booster pu=ps.
These pc=ps take suction frca either of
!.e redundant nuclear services river water circuits and discharge through the cooling coils to the river.
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6.1-3 A=.
c2 tl-L-73)
3D.6 Deleted 3D.7 CALCULATED RESPONSE Displace =ents of the reactor building under pressure have been co=puted at several points on the vessel.
The predicated displace =ents at these points are graphically presented on Figures 3D-9 through 3D-17.
Accepted standards on displace ent seasure=ents are based on increasing predicated displace:ents (from analysis) by 20 percent to reflect all variables, including precision of =easurements, analysis techniques and construction variables such as variation in material properties and dimensions.
The configuration of the radial displace =ent of the containment building under internal pressure is shown ec !1gure 3D-9.
The anticipated cracking of the wall and the done a =aj or discontinuties is indicated in Figures 3D-6 through 3D-3.
3D.3 FORMS FCR RECORDING THE MEASURED DISPLACEMENTS, STRAINS AND TE'9 ER.CRE S Typical forms for recording the measured displace =ents, strains and te=peratures are shown on Tables 3D-2 to 3D-5.
3D.9 INTERPRETATICN OF RESULTS The results of the tests will be interpreted and evaluated af ter co=pletion of the tests by comparison with expected results.
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Table 6.2-11 NPSE REQUIREMINTS FOR THE REACICR SUILDING SPPdY PUMPS AND THE DECAY HEAT RF.MCVAL PR!?S Decay Heat Re= oval Pu=ps Reactor Building Spray Pu=ps Suction Flow NPSH NPSH NPSH Flow NPSH NPSH NPSH Frca gp:
Avail.
Req'd.
Margin gpa Avail.
Req'd.
Margin ft.
ft.
f t.
ft.
ft.
ft.
a a
Borated
- 3000 117.4 7.0 110.4 1500 117.4 12.0 105.4 b
b Water 64.4 57.4 68.4 56.4 Storage Tank Design C
3000 21.9 '
7.0 14.9 1500 9'
12.0 9.9 Reactor
- Bldg.
Runoutd e
3 Su=p 4150 17.87' 13.24 4.63 1S70 18.6' 18.0 0.6 Corresponds to a level of 53'-5" in Borated Water Storage Tank.
a.
b.
Corresponds to a level of 7'-0" in Borated Water Storage Tank.
c.
Assumes te=porary strainer in suction line to pu=ps has been re=cved.
d.
Maximum calculated DH flow occurs with a postulated break in the core flooding line between check valve CF-VSA and the core flood no::le, with flow control valve (DH-V-128A) fully open.
Long term cooling is provided by opening DH-V-7A or 3 to divert a fraction of the discharge of the Decay Heat Removal pu=p to the suction of the Makeup (MU) pu=ps.
The suction require =ents of the MU pumps do not change significantly, and a suction head of 43.2 feet is provided to the MU pu=ps in this mode.
c.
Maxi =um calculated SS flow occurs with flow centrol valve (3S/VlA/B) fully open.
(Runout flow) f.
Minicus NPSH conditions occur at a su=p level of 288.0 feet and su=p water at saturation te=perature of 250 F corresponding to contain=ent pressure of 29.8 psia.
For su=p te=peratures above 212 F, an equilibriu saturation condition is assu=ed between su=p te=perature and contain=ent pressure.
Although the NPSH cal-culation takes credit for contain=ent pressure for su=p te=peratures above 212 F, the intent of Regulatory Guide 1.1 is =et because the saturation pressure used in the calculatica is the =ini=um static pressure which would exist.
Available NPSH for nc=inal f'.ow was calculated usin3 conservative assumptions and applying the mani=um suction line losses to both the Decay Hea; and Suilding Sprav Pucos.
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6.2-68 A=. 62 (1-2L ~3)
6.3.3 PERFOR"ASCE E'... CATION All co=ponents of the reactor coolant syste have been designed and fabricated to ensure their high integrity and thereby =inimize the possibility of their failure.
The reactor coolant systa=, the safety factors used in its design, and the spccial provisions taken in its fabrication to ensura quality are described in Chapter 5.
However, in the event of a reactor coolant syste piping failure, emergency core cooling is provided to ensure that the core will be cooled and will not lose its geo=etric configur-tion. This e=ergency core cooling standby saf ety feature is provided by the cot e flooding syste= and two independent, full-capacity c=crgency core cooling trains of equipment.
The basic design objective of the e=ergency core cooling syste (ECCS) is to termin-ste the te=perature transient and thus =aintain core geo=etry in the event of a loss-of-coolant accident. To evaluate the loss-of-e=ergency accident, anaylses were per-formed for breaks in the hot and cold leg piping. The =axi=u: break size in each pipe was equivalent to the double-ended rupture of that pipe.
Hizh-Pressure Injection Syste:
One HPI (=akeup) pu=p is in operation during nor=al plant conditions, a:,d upon receipt of an engineered safety feature actuation signal, one of the two re=aining HPI pu=ps is actuated. LCCA analyses have shown that the flow fro = one =akeup pu=p is sufficient to prevent core da: age for the s= aller 1:ak ri:es that do not allow the reactor coolant syste= pressure to decrease rapidly to the point at which low pressure injection is in-itiated. The HPI syste= valves are designed to open within 10 seconds af ter receiving an actuation signal. One of the HPI pu=ps is normally in operation, furnishing nor=al
=akeup to the reactor coolant syste=, and a positive static head of water ensures that all pipe lines are filled with coolant. The HPI lines contain thermal sleeves at their connections into the reactor coolant piping to prevent overstressing the pipe juncture.
Operation of this syste= doas not depend on any portion of another engineered safety feature. However, the syste= can be operated in series with the LPI systc= if needed, in the recirculation =cde.
Low-Pressure Injection Svste=
Two LPI (decay heat) pe=ps are actuated upon receipt of a safety features actuation signal. These pumps will deliver the design flow rate for initial inj ection of bor-ated water to the reactor vessel through either of two separate injection lines when the vessel's pressure is approxi=ately 100 psig. LOCA analyses have shown that one of the two LPI flow paths will =sintain core cooling.
Following a LOCA, assuming a si=ultaneous loss of nor=al power sources, the design of the emergency power source and the LPI syste= ensures full operation within sufficient time to =aintain core cooling.
The LPI syste= is connected with the ECCS in three respects:
(1) the RPI and the LPI systems share co==on suction piping fro = the EiST; (2) the LPI syste= and the core flooding syste utilize co==on injection no :les on the reactor kJ4 '# [ }/
A=.
62
- 1-aa-7si 3,3_12
TABI.E 6.3-4 PROCESS INFORMATION Emergency condition Test condition flode Pressure, Temperature,
- Flow, P res su re,
Temperature,
- Flow, Ihh_.
_l"O U _ __ _ _ _ _ F rpm Comments psig F
gym Ccmments 1
30 80 (40 min) 6050-5000 (2) 30 80 (40 min) 3000/1500 1 Dit/BS pump oper 2
30 80 (40 min) 5550-5000 (2) 30 80 (40 min) 3000/1500 1 Dil/BS pump oper 3
36-atm 240-120 4500(500)(4 (3)
NA NA 4
30 80 (40 min) 1800-1500 (2) 30 80 (40 min)
NA 1 US pump oper 5) 36-atm 240-120 1500 (3)
NA NA NA 30 80 (40 min) 3750-3000 (2) 30 80 (40 min) 3000 1 Dit rump oper 36-atm 240-120 3000(700)
(3)
NA NA NA 6
180 80 (40 min) 3750-3000 (2) 180 80 (40 min) 3000 1 011 pump oper 186-150 180-120 3000(700)
(3)
NA NA NA OT Y
7 NA NA NA 30 80 (40 min) 3000 1 Dit pump oper E
8 100 80 (40 min) 3750-3000 (2)
NA NA NA LOO-atm 180-124 3000 (3)
NA NA NA
/* n 49 30 80 (40 min) 500 (2)
NA NA NA b0 1d6-150 240-120 500 (3)
NA NA NA
}
P h
Wll 1000 80 (40 min) 500 (2)
NA NA NA Tested duting
.1000 240-120 500 (3)
NA flA NA normal operation b
for partial flow 12 600 240-120 250 (3)
NA NA NA 600 80 (40 min) 250 (2)
NA NA NA 13 600-0 Ril ambient Tank empty NA NA NA Tested for opening Q3 in 430 s of check valves
! 3 during cooldown fotes:
(1) All pressures, flow and temperatures shown are estimated values.
(2)
Suction from the BUST.
(3)
Suction from the sump.
( ', )
Used only for small breaks where recirculation is through IIPI pumps.
(5)
Flow is limited by test ]ine size.
O 7.3.2.2.5 Accident Monitoring Instru entation Monitoring instru=entation in the control roo= allows the operator to:
1.
recognize that an accident condition exists in the plant, 2.
recogni:e the type of accident that has occurred, 3.
=anitor the status of =itigating equip =ent, and 4.
deter =ine that key plant conditions are being =aintained within design basis conditions.
Table 7.5-2 lists the insttu=entation which the operator has available to perfor= each of cl.e above =entioned tasks. Additionally, it lists the type of accidents f or which the instrc=ent must perf or= the task.
Section 12.1.4 describes the radiation =onitoring equipment while Sections 3.10 and 3.11 discuss the seis=ic and environ = ental qualification. Also, refer to Table 7.4-2 for a discussion cf instru=entation required for safe shutdown. Table 7.5-1, in conjunction with Table 7.5-2, provides pertinent inf or=ation such as nu=ber of channels available, indicator rang 3s, accuracies, and type of read-out.
To ensure that the integrity of instru=ents within containment required for post-LOCA =caitoring is =aintained during accident enditions, pressure seals have been prov.ded at conduit ends of all safety related instru=ents required for post-LOCA =onitoring.
(54 'T I'(
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7.5-6a A=,
62 (1_T4_73
b.
7;he line to Jackson fc11cws an entirely different route than the lines to Middletown Junctica. The lines are in physical proxi=ity to each other in the sta: ion switchyard. Nevertheless, the gec=etry of the tcwers =akes si=ul:aneous s:ructural da= age unlikely.
The envi-ronmen:al design of the towers =akes it unlikely that they could both fail from a co==an enviren= ental cause. An environ = ental effect severe enough to over: urn ene tower into the other wculd al=ost certainly have first resulted in loss of off;ite power.
The Middle:own Junctica substation is only 1.5 =iles away, thus c.
reducing line exposure.
d.
An autocransfor:er tie is provided to tie the 500 kv substa: ion, into which Unit 2's generator discharges 1:s cutput, and the 230 kv sub-s tation f ro= which Unit 2's auxiliaries obt in chair power.
Loss of the tie will not shut down :he unit's auxiliaries.
The breaker-and-a-half switching arrange =ent in the 230 kv substation e.
includes two full capacity =ain buses.
Primary and backup relaying has been provided for each circuit along with circuit breaker failure backup switching.
These provisions per:1: the following:
1.
Any circuit can be switched under nor=al at rault conditicas without loss of external power sources.
2.
Any single circuit breaker can be isolated for =aintenance without interrupting the power or protection to any circuit.
3.
Short circuit of a single =ain bus will be isolated without interrupting service to any outgoing line or to the plant for
= ore than 6 cycles.
4.
Short circuit failure of the tie breaker will result in the loss of its two adjacent circuits until it is isolated by disconnect switches.
5.
Short circui: failure of a bus side breaker will result in the loss of onc circuit and one auxiliary transfor=er until it is isolated by disconnec: switches.
6.
Circuit protection will be insured frc= failure to the primary protective relaying by backup relaying.
f.
Transient stability studies for the bulk pcwer trans=issicn syste have been perfor=ed as discussed in 8.2.2.
With the above protective features, :he prchability of loss of core than one scurce of 220 kv power f rc= f aults is icv.
In the unilkely event of loss of all the 230 kv of fsite power, the engineered safety feature loacs will be supplied frc= one or = ore of the re=aining sources of onsite power as dis-cussed in S.3.1.
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3.
.1.1.2 Inspection and Testing of Electric Power System The offsite power system is designed to permit appropriate periodic inspection and testing of inportant areas and features, such as wiring, insulation, connections, relaying, circuit breakers, transformers and switchboards, to assess the continuit:' of the systems and the condition of their cecponents, as well as their ability to perform their intended functions.
I'.:e 230 kv circuit breakers can be inspected, maintained, and tested as follows:
a.
The 230 kv transmission line circuit breakers can be tected en a routine basis. This can be accomplished on the breaker-and-a-half scheme without removing the transmission line f rom service.
b.
The 230 kv generator circuit breakers of Unit 1 can be tested with the generators in service since two breakers, each fully rated, are provided with the generatcr.
Trans=ission line protective relaying can be tested, maintained, and checked for auto:stic transfer during plant shutdown. Design capability exists for performing this test during plant operation, as required to demonstrate system perfor=ance and component integrity. The ability to transfer the power circuits within 6 cycles mitigates system disturbances sud insures ccntinuity of operation.
Eb0 '# ? 1 An-62 fi-L-73) 3,3_33
8.2.1.1.3 AEC Safety Guides 6 and 9 These guides apply prizarily to the onsite pcwer system and are discussed in 8.3.1.
Compliance with these standards is assured by the provision of two independent sources of prefered power from the transmission network with essentiaily instant availability.
3.2.2 ANALYSIS (07FSITE PCh'ER)
Steady state and transient stability analysis have been cade to determine the perfor=ance of Three Mile Island 2 as well as the transmission network during contingency situations.
The results of these studies have shown that no unit instability, system instability, trans=1ssion line overload or cascading outages will occur as a result of a 3-phase fault and outage of any transmission line e=anating from either the Three Mile Island 230 kv or 300 kv bus.
The planned transmission syste seets the Mid Atlantic Area Council's " Reliability Principles and Standards" and has been approved bf the Council.
In addition studies have shown that the sudden loss of the output of Three Mile Island 2 by itself or along with the next largest unit (Three Mile Island 1) will not ratuit in any syste or unit instability, transmission overloads, cascading outages er intolerable voltage conditions in the network.
The Three Mile Island busses are interconnected to the Pennsylvania-New Jersey-Maryland Interconnection which is also interconnected with other power pools.
This network has proven to be of high reliability with only one interruption of the 2'O kv network occurring which resulted in loss of generation and load. This interruptivn occurred on June 5, 1967 and was caused by the tripping of an overloaded circuit and loss of generation, These outages caused cascading tripping which resulted in the loss of the eastern portion of the PJM inter-connection.
However, the 230 kv network 1 =ediate to Three Mile Island remained in service.
The netwock was restored in five hours.
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/m. 62 (1-2h-73)
S.2-3
The resin addition tank as shown on Figure 11.2.1 is si:ed to per=it gravity replace =ent of resin in any of the de=ineralizers of the =akeup and purifi-cation, radioactive liquid waste, or spent fuel purification syste=s.
The routing of resin to the proper pair of de=ineralizers is assured by the =anual installation of a flexible connection between the resin tank outlet and the resin fill line to the de=ineralizers being serviced.
The hydrogen supply syste= consists of a hydrogen manifold which =aintains adequate hydrogen partial pressure in the gas space of the =akeup tank to li=it the dissolved oxygen content in the reactor coolant to an acceptable level during nor=al operation. Radundant isolation devices will be provided in the hydrogen charging line, with safety grade control for valves and position indication.
Two nitrogen =anifolds are provided. One supplies fresh nitrogen blanket gas to the =akeup tank, the low pressure vent header portion of the radio-active waste gas disposal systa= and provide =iscellaneous nitrogen source throughout the nuclear syste=s.
The second nitrogen =anifold provides high pressure nitrogen to the core flooding tanks.
9.3.4.2.2
.Syste= Description The chemical addition syste= is co= prised of a nu=ber of individual syste=s (each having their =2jor equip =ent components and piping syste= with valves and contrels) as required to perfor= various functions related to the operation of the pri=ary syste=, the spent fuel pools, and the radioactive liquid and gas waste disposal systems. The che=ical addition syste= is shown in Figure 9.3-7.
The =ajor equip =ent components of the che=ical addition and sampling syste=s are listed with the pertinent data in Table 9.3-6.
Other chc=ical addition equip =ent is provided in the turbine building as re-quired to =aintain the quality of feedwater to the steam generators as indi-cated in Table 10.4-1, i
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9.3-27 J.:. 62 s1-2L-73)
The ioni:ation type duct s=oke detectors provided are designed to ensure unifor= sensitivity in air velocities ranging fro a low of 500 ft/=in to as high as 3100 ft/=in.
The detector, therefore, does not require co=pensat-ing adjust =ents f or operation st specific air velocities.
Co=binctica rate of rise and fixed te=perature detectors provided throughout the units are rated at either 135 ? or 190 F.
They also have rate of rise features which respond when the air '.s heated at a rate greater than 15 F/=in.
Flace detectors provided in the unit respond directly to the presence of flace.
This type of detector is e= ployed in the diesel generater roo=s and in the fire pu=p house because oil fire in these roo=s =ay develop rapidly.
l Fla=e detectors sense the =odulated radiation in the infrared region e=anating fro = flaces. The radiatica =ust be sustained for at leas 10 seconds for the detector to respond.
Sensitivity changes in the fla=e detector are negligible in a te=perature range of -20 ? to 175 F.
Relative Hu=idity up to 807. continucus and 90% inter =ittent will not appreciably affect sensitivity.
9.5.1.4 Tests and Inspections In accordance with NFPA No. 69 " Explosion Prevention Syste=s", the halon syste=s for the Air Intake Tunnel will be inspected and tested every three conths.
Following each inspection and test a report will be prepared describing the condition of each co=ponent and wiring that has been integrated into the syste=.
Resistance measure =ents of all wiring will be =ade to deter =ine in-sulation breakdown, ground conditions that could be affected by =oisture, etc.
The capability of each detector used in the syste= will be checked regularly.
Resistance measurements will be performed on all wiring that electrically connects each detector to the centrol panel.
In accordance with hTPA No.13 " Sprinkler Syste=s, Installa:icn" and with NFPA No. 15 " Water Spray Fixed Systa=s" all deluge, sprinkler and preaction syste=s will be inspected and tested annually.
Following each inspectica and test, a report will be prepared describing the condition of the syste=.
A visual inspection is made of the detactor, sprinkler and piping syste=.
The =ain gate valve will then be closed and the =ain drain valve opened.
One heat-actuated deviced will be operated by an electrical test set or hot war.er (175 F or more) in a container which can be placed around the heat actuated device so as to submerge the bulb.
This will open the solenoid valve which in turn releases the clapper latch holding the deluge valve closed. After operatica of the deluge system, each valve will be reset.
The re=aining heat actua:ed devices will be tested to see that the solencid valve operates.
The break glass =anual control station will be tested by opening the cover and depressing the interior cperating lever.
Water will drain frc= the drain nipple.
The syste will then be put back into operation.
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62 (1-2--73}
T.o a x 13 x 10 inch valves set at 1050 psii One 3 x 10 x 10 inch valve set at 1C60 psig One 8 x 10 x 10 inch valve s e t a t 10~ 0 ps:.g
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Cne 3 x 10 x 10 inch valve set at 1030 : sir Cne. 8 x 10 x 10 inch valve set at 1102'psig The maximum pressure drep from the steam generatcr to the first safety valve set at 1050 psig is 22 psi and actual selected valve capacity is 55 greater than desien.
10.3.3 EVALUATICd The following components of the Main Steam Supply System are protected against aircraft impingement and are also designed in ccmpliance with d e seismic Class I requirements as discussed in 3.2.1.
a.
All main steam lines between the steam generatcrs and the =ain and branch steam isolation valves.
b.
Main and branch steam isclatien valves.
c.
At=cspheric du=p valves.
Since there are no open crcss-connections between steam generators, the rupture of a line frca cne steam generatcr will not blow the other steam generator dry and also insures that a steam supply line is available to the main steam g
generater feedpu=p turbines and the emergency steam generator feedpump turbine.
(
The function of the main and branch steam isolatien valves is to maintain cen-tain=ent integrity in the event of a steam line rupture wisin the containment.
Safety relief valves are located upstream of the isolation valves and will dissipate all the energy existent or produced in the ':cclear Steam Supply System to the environment when required.
In addition, upstream of the rain steam isolation valves are takeoffs which serve as steam bypass to the cen-denser, and centrolled steam relief to the enviren=ent. See 10.4.4 for additional informaticn en the Turbine Bypass System.
10.3.4-
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.x The valves and =ajcr cc=penents of the Main Steam Supply System are subjected to =anufacturer's shop tests including hydrcstatic and perfcr=ance tests.
Check analysis for de ASTM A333 Gr. 6 piping material within the system is made to assu e confor=ance v.c the piping =aterial chemical requirener:ts.
Tests and inspecticn of the Main Steam Supply System piping, such as the radio-graphic inspecticr. for the welds and the hydrcstatic leak test prior to initial cperation is =ade in accordance with the requirements of the ccd:s and stand-ards to which de system is designed.
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Describe the =cdes of failure :hs are considered in the design of the spent fuel cask crane and reactor pola:
Jane such as breaking of cables, lifting slings, sheared shafts, keys, stripped gear teeth, and brake failures.
- Also, discuss the li=ita:1:ns and control that will exist in handling cbjects over an opened reactor vessel.
RISPONSE
!.lchcugh no sys:ematic f ailure : de analysis was performed for the design of the Reactor Building polar crane, or the fuel handling crane, the poten:ial for f ailures in the braking and hoisting systa=s was considered.
As a result, the sa:e:y features described in FSAR Section 9.1.4.3 have been incorporated in the main hoists of the Reactor Building pclar crane and the fuel handling crane.
The handling of Icads Over an opened rea ::r vessel vill be administratively centrolled in aca^-d with approved precedures.
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. yplement 2 02.31/10.4.5 scheduled surveillance or =aintenance which would require access through these doors, najor corrective maintenance and surveillance of the doors the=selves are the only conditions requiring door openings.
To minimize the possibility of flooding the Control Building Area as a result of a circulating water rupture in the Turbine Building, the following restrictions apply to operation of the watertight doors between the Turbine Building and Control Build-ing Areas at the 281'6" level when the circulating water system is in operation:
a) All three watertight doors will be shut and locked with keys under the adminis-trative control of the unit superintendant.
b)
If access through the water tight doors is required for =aintenance or surveil-lance
- 1) Only one watertight door will be opened at a time.
- 2) The work will be planned and executed such that the door is open only when necessary for the transport of equipcent and material for the mini:um time practical. The door will be shut and dogged when not required for access.
- 3) Work requiring periodic openings of watertight doors shall be cc=pleted in less than 48 consecutive hours.
- 4) During periods wher. the door is open, the hatch will be kept as free as possible from obstructions which could foul the door and prevent its closing.
- 5) During periods when a watertight door is unlocked, an attendant will be stationed to operate the door in accordance with these requirements.
Should a circulating water rupture occur during the period when a water-tight door is open, the attendant will inform the control room to stop the circulating water pu=ps and atte=pt to close the watertight door.
If the circulating water system is secured (all 6 pumps off and their discharge valves shut) the watertight doors cay be unlocked and opened without observing the above restrictions since there is insufficient residual water in the circulat-ing water system to flood into the Control Building Area.
A control room annunciator will be provided for all three watertight doors to indi-cate when any of the doors is not fully shut.
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S2-1272 An. 62 (1-Zh-73)
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Supple ent 3 222.41/(None)
(Continued) t Fire stops are provided to prevent the propagation of fire along a cable tray.
Fire stops are provided in vertical cable trays only when the tray passes through penetrations in concrete floors.
Fire stops are provided in heri: ental cable trays only when the tray passes through penetrations in concrete valls.
Quality Control require:ents for the installation of fire stops are given in the Design Specifications. The Quality Control requircrent is that the constructor perform an installation inspection to verify that the =aterials used to seal the penetration are installed in accordance with the A-E drawings.
No testing of fire stops is required after installation.
Sealing of all penetrations in the walls and floor of the Cable Spreading Roon will be done as discussed above.
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