ML20237A115

FSAR Section 5, Containment

ML20237A115
Person / Time
Site:	Pilgrim
Issue date:	11/05/1987
From:	BOSTON EDISON CO.
To:
References
CON-#487-5061 2.206, NUDOCS 8712140269
Download: ML20237A115 (60)

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SUMMARY

DESCRIPTION ywr u 00CACW i ' ~i 5'.1.1. General NC 'f i The containment systems of Pilgrim Nuclear Power Station' utilize a l "multibarrier* concept which consists of two systems. The Primary Containment System (PCS) is a pressure suppression system which'iorms f .the first barrier. 'The Secondary containment System (SCS) is a system, which minimizes ~ the grot.nd level release of e.irborne . radioactive materials, end for,ns the second barrier. The fcel, fuel. j

cladding, and Reactor Primary System (RPS) form additiona'i barriers j

~ to the release of fission products and are described in Section 3.2. .i 5.1.2 Primary Containment System The PCS houses the reactor vessel,'th'e Reactor Colant Recirculation ~ j System and other branch connections of the Reactor Coolant System (RCS).. The primary containment. is a pressure suppression system 'l cer.sisting of a drywe13, pressure suppression chamber which stores a l large vclume of water. a connecting vent system between the drywell C,# and water p u l, isolation valves, vacuum relief I system,- containment ctoling systems, and other service equipment..The drywell.is a steel j ' pressure ' vessel in. the shape of a light bulb, and the -pressure .] suppression chamber is a torus shaped steel pressure vessel located ..a below and encircling; the drywell. the PCS.is designed te withstand the forces from any size breach of the. nuclear system primary barrier up to and including an instanteedus circumferential-' break of the reactor recirculation piping, and provides a holdup time for decay of.any radioactive materi'1 released. The PCS also stores sufficient water to condente d { .the stes.m released as a result cf a breach in-the nuclear system primary barrier and to supply the Core Standby Cooling Systems ] (CSCS). 1 i 5.1.3-Secondary containment System J The SCS encloses the PCS, the refueling and reactor servicing areas, new and spent fuel storage facilities, and other reactor auxiliary systems; The 505 serves,as the only containment during reactor refueling and maintenance operations, when the primary containment is

open, and as an additional barrier when the I.'S is functional. The i

SCS consists of the reactor bui2. ding, Standby Gas Treatment System j (SGTS), main stack, Reactor Building Isolation and Control System i -(RBICS), and other service equipment. j l i .( 8712140269 871105 DR ADOCK 05000293 j PDR G.1-1 4 - - ~ ~ -

i t r y y (;. .~t. ib. n? jy O ...r PNPS-FSAR i [ "t The' SCS <is ' designed to withstand the e ximum postulated seismic Jelehnand be capable of providing holdup treatment,ipoint for any fission produc i

event, and an elevated l

,i [. 9a the RwJtor Building is designed to provide protection for' the 1 l - engineered.safeguarcs and nuclear safety. systems located in the - building. from 'all postulated environmental events including tornadoes. l N, s. l ~- t p ~ '*, l 1 l f 'I. \\ w. m b 1 i l r e .{ '(' i 5.1-2 L ~

, y> m el ' ( PNPS-F2AR 5.2 PPIMARY CONTAINMENT SYSTEM 5.2.1 Safsty objective The cafety objective of the Primary Containment System (PCS) is to j provide the capability in conjunction with other gefeguard features, to limit-the release of fission products in the event of a postulated i design basis accident (DBA) so that offsite doses sculd not reeed the guideline ~ values set torth in 10CFR90. 5.2.2. Safety Design Basis 1. The PC'3 shall have the capability of withstanding.the conditions j which could result from.hy of the postulated DBAs for which the PCS is assumed to be fuaetional, iraluding the largest ameunt of energy relesse and mu s flow associated with the accident. ,Y 2. The PCS thall have a margin for 2netal water reestions and'other chemical reactions subraquent to any postulated DBA for which the FCS is assumed to be functional, consistent with the perfo6 nance obj ec-t.ives of the nuchar safety systems and engineered j L afeifuards. 3. The PCS shall have the capability to maintain its functional g integrity during any postulated txternal er environmental event. 4. The PCS shall have the capability to be filled with water as an accident recovery method for any postulated DBA in which a breach of the nuclear system primary barrier cannet be sealed. 5. The PCS, in conj unction with other Nuclear Safety Systems ar,d engineered safeguards, shall have the capability to limit leskege ~ durdng any of the postulated Dr.,As for ebich it is assumed to be funct-lonal, such that offsite doses do not exceed the guideline valu&s set forth in 16CTR100. 6. The PCS shall have the capability to rapidly isolate all pipes or duett, necessary to establish the primary containment barrier. 7. Thb PCS shall have *,he capability to stcre sufficient ws.ter ta supply the Core Standby cooling System (CSCS) requirements. 8. The primary containment shall have the capability to be meintained during normal operation within the range of initial conditions assumed in the Station Safety Analysis in Section 14. 5.2.3 Description j l 5.2.L1 General The design employs a I.cw Leakage Pressure Suppression Containment l J System which houses the reactor vessel, the reactor coolant recirculating, loop s, and othsr branch connections of the Reactor ) Primary System. The Pressure Suppression System consists of a j l 5.2-1 ~_

y I PNPS-FSAR =

dryvell, a pressure suppression chamber (torus) which steres a large volwae of water, a cennecting vent system between the drywell and the pressure suppression pool, isolation valves, Vacuum Relief System,

.l containment cooling Systems, and other service equipment. The. PCs design parameters are given on Table 5.2-1. ~ In the event of a Process System piping failure within.the drywell, reactor water and steam will be released into the drywell gas space. The resulting increated drywell pressure forces a mixture of air, steam, and water through the vent ; system into the pressure suppression pool. ~The steam condenses rapidIV in the supprensich pool resulting in rapid pressure reduction in the drytell.- Air transferred. during -reactor blowdown-to the suppression chamber pressurizes the chamber, and subsequently is vented to. the drywell l through the vacuum relief system as th'e preisure in the drywell drops lg below that in the suppression chamber. Cooling systems are.provided to remove heat from the water in the i suppression chamber. This provides for continuous cooling of the pr.in..ry centaiuent under the postulated DBA conditions for which the PCS is assumed to be functional. Isolation valves are provided to O ensure containment of radioactive materials within the primary containment,. which might be released from the reactor to the i - - containment doring the course of an accident. Other service equipment is provided to maintain the containment within its design parame' era during normal ope; ation. The drywell (primary containment) coolers are designed to maintain drywell atw sphere temperatures within an acceptable range dyring normal station operation. See Table 5.2-2. The reduction of atmosphere temperature by the coolers will also result in partial condensation of water vapor when the incoming humidity levels are hich. The drywell fan meters are rated for continuous operation in atmospheres having 100 percent Eh and 100F temperatures. In the design ef the cooler, the motor has been placed in the exhaust of the cooler where the leaving air temperature is a maximum of 95'F, so that the motor it exposed to the lowest humidity and lowest tuaperature atmosphere available within the drywell. Pressure increases to the 2.5 psig high drywell pressure condition used to sense a possible loss of coolant accident (LOCA) would not affect the continued operability of the coolers. The drywell coolers are automatically shut down in the event of a LOCA combined with the less of offsite ac power. The drywell coolers, including the fans, with their pcNer and control systems were tested during the preeperational tests at the station to demonstrate the required operability of the power and control systems, the fahs, and the pteactor Building closed cooling bater supply to the coolers. The capability of the coolers to maintain the required drywell atmosphere temperatures was verifidd during the ) startup program as the drywell heat loads increased during the heatup and pressurization of tha Nuclear Steam Supply System. 5.2-2

~ ~ ~ N PNPS-FSAE The Reactor Building Closed Copling Water System (RBCCWS) piping supplying the drywell coolert will be revised to seistaic Class 2 to

1. tintain the pressure boundary integrity of this piping under seismic loading.

Refer to Section 10.5.5.1. The drywell cooler? were originally purchased as seistnic c' tis I equipment to serve as pressure boundary only. The PC$ design loading considerations are given in Section 12 and i Appendix C. The Station Safety Analysis presented in Section 14 demonstrates the effectiveness of the PCS as a radiological barrier. In addition, primary containment pressure and temperature transients from postulated DBAs are also presented in Section 14. 5.2.3.2 Drywell The drywell is a steel pressure vessel with a spherical lower t portion, 64 ft in diameter, and a cylindrical upper portion 34 ft 2 inches in diameter. The overall height is approximately 110 f t. design, fabrication, inspection, and testing of The the drywell vessel complies with requirements of the ASME Boiler & Pressure Veitsel Code Section III, Subse6 tion B, R2quirements for Class e B Vessels, which. pertain to containment vessels for nuclect power stations. The drywell is designed for an internal pressure of 56 psig-coincident with a temperature of 281*F vith applicable dead,

live, and seismic loads imposed on the shell.

Thus, in accordance with the ASME Code, Section III, code Case N-1312-(2), the maximum drywell presser.t is 62 psig. . Thermal stresses in the steel shell due to temperature grndients are taken into account in the design. Special precautions not required by codes were taken in the fabrication of the steel dryvell shell. Cha rpy V-notch specimens were used for imnact testing of plate and forging material to give aesurance of proper agterial properties. Plates, forgings, and pipe associated with the drywell have an initial HDT temperature of O'F or lower wher. tested in accordance with the appropriate code for the materials. It is intended that the drywell will not be pressurized or subjected to substantial stress at tempera +.ures below 30'F. The drywell is enclosed in reinforced concrete for shielding purposes, and to provide additional resistance to deformation and buckling in areas where the concrete backs up the steel shell. Above the transition zone, the drywell is separated from the reinforced concrete by a gap of approximately 2 in. Shielding over tFe top of the drywell is provided by removeable, segmented, reinforced concrete shield plugs. In additien to the drywell head, one double door air lock and two bolted equipment hatches are provided for access to the dryw, ell. locking mechanisms on each air lock door ere designed so that a tight The seal vill be maintained when the doors are subjected to design ( pressures. The doors are mechanically interlocked so that neither door may be operated unless the other door is closed and locked. The 5.2-3

n PNPS-FSAR drywell. head and equipment hatch covers are bolted in place and sealed with gaskets. The spectrum of crimary system leak rates up to a double ended blowdown of a recirculation line has been analyzed relative to the temperature and pressure response of the drywell. Steam issuing-from a leak and exp nding at constant enthalpy may result in a superheated i. containmer.t atmosphere. The maximum amount of superheat possible is a function of both the source pressure (reactor pressure) and the receiver pressure (drycell). The enthalpy of saturated steam goes through a maximuin value at a reactor pressure of 400 to 500 psia. Steam issuing from a leak at this pressure will result in the maximum superheat for a given containment pressure. If a steam leak occurs, the containment pressure and temperature increase at a rate depender.t on the size of the leak, containment I characteristics, and the pressure of the reactor. The containment pressure and temperature rises as noncondensable gases are swept into the suppressi9n chamber. Containtnent pressure levels off after all anacondensable gases are driven into the suppression chamber. The l containment shell temperature rises as steam condenses on the relati' sly cool wall. When the drywell shell temperature reaches the v saturation temperature dictated by, this centsinment pressure, steam - _ condensation is terminated. The only energy available to further increase the wall temperature is the superheat energy. The result is a decrease in the rate of temperature rise of the drywell shell and ) an increast in the bulk atmosphere temperature of the drywell. Figure 5.2-1 illustrates the reactor vessel pressure response to steam leaks ranging in size from 0,02 to 0.50 fta. Figures 5.2-2 through 5.2-6 illusto te the containment response to steam leaks covering the same s2te range. The time it takes for the containment ) wall to reach 281'F as. a function of leak size is shown on Figure 5.2-7. For the complete vectrum of steam leak sizes, the I time for the containment wall to reach 281'F is always greater than f 2,000 sec, The responss of the containment to small steam leaks is slow, but the continued high reactor pressure results in high containment temperature, given enough tica. Leaks so small that the high drywell pressure trip does net occur will not result in a high temperature. Leaks large enough tc result in a high containment temperature will be large enough to sweep air into the suppression chamber and result in significant drywell pressure increase. Large leeks will either depressuriz> the reactor rapidly or result in auto-relief such that steam temperatures above 281*F do not pe7sist long enough to b( of concern. Figure 5.2-7 shows the steam line breat. spectrum of interest vs time to reach a wall temperature of 281'F, which in all cases is in excess of 2,000 sec. The break sizek for which temperatures higher than 281'F steam temperature are possible for sufficient durations to affect the wall temperature, are of such a size that the opernor can take manual action within 30 min 'fter I a t r>

incident, and prevent the containment wall from ever exceeding i

281'F. I ) 5.2-4

PNP 3-FSAR 4 Activation of one of the two containment sprays any time before the wall temperature reaches 281*F would be effective in terminating the temperature rise because the superheat is quickly removed. The spray nozzles are designed to give a small particle size, and the heat t:ansfer to the subcGoled spray is very effective. Since the total amount of heat in the drywell atmosphere is low relative to the spray rate, the containment atmosphere temperature is quickly reduced to n.est the spray temperature. l A drywell pressure condition exceeding 10 psig was selected as the basis for determining when to initiate the containment spray. See Figure 5.2-8 for time required to reach 10 psig. Based on the results of Figure 5.2-7, the operator will be instructed to initiate the containment sprals if containment pressure exceeds 10 psig for longer than 30 min. This procedure will ensure that the containment wall never exceeds 281*F. Depressurization of the reactor vessel can take place at the normal rate, but depressurization is not required ,i to ensure that the wall temperature remains below 281*F. The environmental conditions considered in the design of the reactor protective system instrumentation, engineered safety feature equipment, and de qualification tests that have been conducted are described in Section 7.1.8. 5.2.3.3 Pressure Suppression Chamber and Vent System 5.2.3.3.1 General

'l 1

The pressure suppression pool, which is contained in the pressure suppression chamber, initially serves as the heat sink for any postulated transient or accident condition in which the normal heat sink, main condenser, or Shutdown Cooling System is unavailable. Energy is transferred to the pressure suppression pool by either the ~ discharge piping from the reactor pressure relief valves or the Drywell Vent System. The relief valve discharge piping is used as the energy transfer path for any condition which requires the operation of the relief valves. The Drywell Vent System is the energy transfer path for all energy releases to the drywell. Of all the postulated transient and accident conditions, the instantaneous circumferential rupture of the reactor coolant recirculation piping represents the most rapid energy addition to the pool. For this accident the vent system, which connects the drywell and suppression chamber, conducts flow from the drywell to the suppression chamber without excessive resistance and distributes this flow effectively and uniformly in the pool. The pressure suppression pool receives this flow, condenses the steam portion of this flow, tnd releases the noncondensable gases and any fission products to the precsure suppression chamber air space. 5.2.3.3.2 Pressure Suppression Chamber The pessure suppression chamber i t, a eteel prassure vessel in the shape of a torus below and encircling the drywell, with a centerline vertical die of 29 ft ( an and a hacizontal die of 131 f t 6 in. The 5.2-5 _m___

i FNPS-FSAR } pressure suppression chamber contains approximately 84.000 ft' of water and has a net air space above the water pool of approximately 120,600 f t'. The suppression chamber will transmit seismic loading l to the reinforca$ concrete foundation slab of the Reactor Building. Space is provided catside of the chamber for inspection. The toroidal suppression chambor is designed to the same material and code requirements as the steel drywell vessel. The material has an NDT temperature of 0*F or bsse 5.2.'3. 3. 3 Vent System Large vent pipes connect the drywell arm the pressure suppression chamber. A total of eight circular vent pipes are provided, each having a dia of 6.75 ft. The vent pipes are d6 signed for the same pressure ed ' temperature conditions as the drywell and suppression chamber. Jct deflectors are provided in the drywell. at the entrance of each vent pf pe to prevent possible damage to the vent pipes from jet forces which might accompany a pipe break in the drywell. The vent pipes are fabrica,ted of SA*516

steel, and comply with requirements of the ASME Boiler and Pressure Vessel Code, Section III, Subsection B.

The vent pipes are provided with expansion joints which are enclosed within sleeveu, to acenmodate differential motion - -between the dryrell and suppression chamber, n The drywell vents are connected to a 4 ft 9 in dia vent header in the I form of a turus whien is contained within the cirspace of the suppression chamber. Projecting downward from the header are 96 downcomer pipes, 24 inches in dia, terminating approximately 3.00 to 3,25 f t below the water surface of the pool. The vent header haa the same temperature and pressuro design requirements as the vent pipes. Vent pipes and vent headers are braced to withstand expected loads fcom steam blowdown into the pool. 5.2.3.3.4 Pressure Suppression Pool The pressure suppression pool is approximately 84,000 ft' of demineralized we t e '- contained within the pressure suppression chamber. It serves both as a heat sink for postulated transients and accidents and as a source of water for the CSCS The ' suppression pool receives energy in the form of steam and water from the reactor pressure relief valve discharge piping, or the drywell vent system dowrcomers which discharge under water. The steam is condensed by the suppression pool. The condensed eteam and any water carryover cause an increase in pool volume and temperature. Energy can be removed from the suppression pool when the Resitlual Hcat Removal System (RHRS) is operating in the suppression pool cooling mode. Th suppression pool is the primary source of water for the Core Spuy and Lov Pressure Coolant Injection (LPCI) Systems, and the secondary source of water for the Reactor Core Isolation Cooling (RCIC) and High prersure Coolant Injection (HPCI) Systems. The water

l 'e 1 PNPS-FSAR i j E level and temperature of the suppression pool are continuously mo'nitored in the main control' room. ~ 5.2.3.4 Penetrations 5.2.3.4.1. General Con'tainment penetrations have the following design characteristics: 1. They are designed for the rsme pressure and temperature conditions as the :!rywell and pressu,re suppression chamber 2. They are capable of withstanding the forces caused by impingement of the fluid from the rupture of the largest i local pipe or connection without failure 1-3. They are espable of accommodating the thermal and mechanical ~ stresses, which may be encountered during all modes of operation including environmental. events, without failtre 4. They are capable of withstanding the maximum reaction that the pipe to which they are attached is capable of exerting The penetration schedule, including the number and size of these ~' penetrations, is shown on Table 5.2-3. Lead combinations and allowable stresses are described in Appendix C. 5.2.3.4.2 Pipe Penetrations Two general types of pipe penetrations are provided. Type 1 is used where the design must accommodate thermal movement. Figure 5.2-9 is typical of this type of penetration. Type 2 is used where stresses + due to' thermal movement are relatively small. Typical penetrations of this-type are illustrated on Figures 5.2-10 and 5.2-11. Figure 5.2-12 shows a typical instrument penetration. The piping penetrations which have special provisions for thermal movement, such as the steam lines, are shown on Figure 5.2-9. In these penetrations, the process line is enclosed in a guard pipe that do attached to the main steam line through a multiple head fitting. This fitting is a one-piece forging with integral flues or nozzles and is designed to meet all requirements of the ASME Boiler and code.l is Pre s s.ure Vessel Code, Section III, Class '3. The forging radiographer and ultrasonically tested as specified by this The guard pipe and 'lued head are designed to the same pressure requirements as the rcess line. The process line penetration sleeve is welded to ae drywell, and extends through the biological shield where it is velded to a two-ply expansion bellows assembly, which in turn is velded to the flued head fitting. The pipe is i guided through pipe supports at the end of the penetration assembly to allow steam pipe movement parallel to the penetration,'and to limit pipe reactions of the penetration to allowable stress levels. t 5.2-7 Revision 2 - July 1983 l

P?iPS-FS AR 1 Ilhere necessary, rta penetr& tion assemblies are anchored outside the containment to limit the movement of the line relative to the containment. The cellows accommodates the relative movement between ' i the pipe and the containment shell.

• ?

The design of the penetration takes into account the stresses associated with normal thermal expansion, live and dead loads, i seismic loads, and loads associated with a LOCA within the drywell. 1 The design takes into account the loadings given above in addition to the jet force loadings resulting from any pipe failure. The resultant stresses in the pipe and penetration for the condition do not exceed 90 percent of the material yi. eld stress. The cold piping, ventilation duct, and instrument line penetrations are generally welded directly to the sleeves. In some cases, where l stress analyses indicate the neeo, double flued head fittings are q -y used. Bellows and guard pipes are not necessary in these designs, ~ since the thermal stresses are small and are accounted for in the design of the weld joint. 5.2.3.4.3 Electrical Penetrations ~ k The electrical penetrations include electrical power, signal, and ) instrument leads. Typical electrical penetrations are shown on ~ ~' Figures 5.2-13, 5.2-14, and 5.2-15. The penetrating sleeve is welded to the primary containment vessel < Medium voltage power penetrations primary seals are made of alumina-ceramic materials. The seals are formed at 1,300*F or higher, and thus the temperatures to which the seals would be exposed during a LOCA would have no adverse effect on their leaktightness : characteristics. The electrical penetrations used for low voltage power, control, and instrumentation cable and for coaxial cable utilize either A1 0 or a i 2 3 bonding resin to maintain thc leaktight integrity of the containment penetrating sleeves. A prototype of the penetration assembly which utilizes a bonding resin has been tested by exposing the interior face of the penetration assembly to the following environmental ec.1ditions : 281*F, 63 psig internal pressure, 90-100 percent rh for 10 days. An additional test at 320*F, 125 psig internal pressure and 90-100 percent rh for 2 hr was conducted. The pressure retaining capability of the penetration assembly was maintained throughout the duration of the tests. The leak rate was monitored during the test and did not exceed 24 cc/hr of nitrogen through the inner seal. The outer seal is not exposed to high temperatures during an accident and therefore the. overall leak rate througn both seals is 10-6 ce/see, Additional tests were planned to certafy the pressure retaining capability of those penetrations utilizing bonding resin at 340'F, 100 percent rh for 30 min. A prototype of the penetrations using a polysulfone seal has been ( qualified to the following environmental conditions: 340*F,110 psig 5.2-8

) PNPS FSAR li for 6 hr: 320'F, 75 psig for 3 hr; and 260'F, 20 psig for 12 days. The inboard and/or outboard seal possessed a leak rate that was less ) than 5.3 x 108 cc/see helium. Section 7.1.8 states that qualification tests were to be conducted on the medium voltage electrical penetrations, including leakage tests, following environmental exposures in excess of the design basis LotA conditions. 5.2.3.4.4 Traversing Incore Probe Penetrations Traversing incere probe (TIP) guide tubes pass from the Reactor Building through the primary containment. Penetration r.f the guide tubes through the primary containment are sealed by mear.s of brazing which meets the requirements of the AStiE Boller and Pressure Vessel Code, Section VIII. These seals would also meet the intent of ( .I Section III of the code even though the code has no provisions for qualifying the procedures or performances. 5.2.3.4.5 Personnel and Equipment Access Locks One personnel access lock is providt' for access to the drywell. The ' lock has two gasketed 6oors in series, and each door is designed to j withstand the drywell design pressure. The doors are rnechanically l interlocked to er.sure that at least one door is locked at all times I when primary containment is required. The locking mechanisms ar2 I y designed so that a tight seal will be maintained when the doors are subjected to either the design internal or external pressure. The seals on this access opening are capable of being tested for leakage. A personnel access hatch Vith tescable sea.is is provided on the drywell head. This hatch 14 belted in place. Two equipment access hatches with testable seals are also provided. These hatches are bolted in place. 5.2.3.4.6 Access to the Pressure suppression Chamber Access to the pressure suppression chamber is provided at two lccations from the Reactor Building. There are two 4 ft dia manhole entrances with double gasketed bolted covers connected to the chamber by 4 ft dia steel pipes. 5.2.3.4.7 Access for Refueling Operations The drywell vessel head is removed during refueling operations. The i head is held in place by bolts and is sealed with a double-seal arrangement. t l 5.2-9 ________w

] ] 1 i i PUPS-FSAR ] 5.2.3.5 Primary Containment Isolation Valves ) 1 5.2.3.5.1 General Criterio The basic function of all primary containment isolation valves is to i provide r,ecessary isolation to the containment in the event of accidents or similar critical conditions when the free release of containment atmosphere cannot be permitted. The containment isolation valves are listed on Table 5.2-4. This table also defines the v'alve status (normally open or normally closed) during normal reactor ope ra tion and shows the signals required to initiate their i desired operation. The primary containment isolation valves are grouped into four basic cla:.ses. ) Class A valves are on process lines that communicate directly with { v the reactor vessel and penetrate the primary containment. These ] lines require two valves in series, one inside the primary containment and one outside the primary containment. They are located as close to the primary containment boundary as practical. Except in the case of check valves, both valves shall close automatically on isolation signal. Both valves shall receive the isola tion (closure) signal even if normally closed during reactor opertcion. Since check valves close on reverse process flow, they j - - are ased to isolate some incoming lines. Testable check valves are 1 used on selected process inflow lines where flow is expected to be i zero, or on lines which have low flow with intermittent use during / normal station operation. All Class A valves except check valves are capable of remote manual control from the control room. Class B valves are on process lines that do not directly communicate with the recctor vessel, but penetrate the primary containment and communicate with the primary containment free space. These lines ~ require two valves, in series, both of them located outside the primary containment, and as close to the primary containment boundary as practical. Except in the case of check valves, both valves close automatically on isolation signal. Both valves receive the isolation closure signal even if normally closed during reactor operation. See Table 5.2-4 for valve status during reactor operation. All Class B valves except check valves are capable of remote manual control from the control room. Class C valves are on process lines that penetrate the primary containment but do not con unicate directly with the reactor vessel, with the primary containment free space, or with the environs. Class C lines require only one valve which cleses automatically by process action (i.e., reverse flow) or by remote manual operation from the control room. ~ Motive power for the valves on process lines which require two valves shall be from physically independent sources to provide a high probability that no single w idental event could interrupt motive poter to both closure devices. 5.2-10

i PNPS-PSAR l l Variations to the above definitions are referent;ed cn Table 5.2-4 by their class designations followed by an "X" suffix. The lines in thi3 class are generally initrument lines or lines used for core cooling. Automatic isolation valves, in the usupt sense, ars not used oa tha inlet lines of the Reae ;cr Core and Containment Cooling Systems and Recctor Feedwater Systems, since opergt O of these systems is essential following a design basis LCCA. Since normal flow of water in these systems is inward to the reactor vessel o: to the primary r,ontainment, check valves located in theep lines will provide automatic isolation, il necessary. No Automatic isolation valves are provided on the Contr.31 Rod Drive System hydraulle lines. These linn see isolated by the normally f closed hydraulic system control valves located in the Reactor .i Building, and by check valves compricing a part of the drive mechanisms. TIP lines and small diameter instrument lines are not provided with automatic isolation valves. 5.2.3.5.2 Additional Considerations Effluent lines such as main steam lines, which connect to the reaetor vessel or which are open to the pritnary containr. ant, have air-powered valves. This arrangement provides a high reliability with respect to functional performnce. These valves are closed automatically by the signals indicated on Table 5.2-4. The MSIV's are also connected to the nitrogen surply system. This redundant source of MSIV actuation results in greattr system reliability. TIP system guide tubes are provided with an isolation valve which closes automatically upon receipt of proper signal and after the 27P cable and fission chamber have been retracted. In series M.th this isolation valve, an additional or backup isolation ~ shear valve is included. Both valves are located outride the drywell. The functi<an of the shear valve is to assure integrity of the containment in the unlikely event that the other isolation valve should fail to close or the chamber drive cable should faal to retract if it should be extended in the guide tube during the time that containment isolation is required. This valve is designed to shear the cable 4.nd seal the guide tube upon an actuation signal. Valve position (full open or full closed) of the automatic closing valves will be indicated in the control room. Each shear talve will be operated independent,1y. The valve is an explosive type valve and each actuating circuit is monitored. In the event of a containment isolation s.gnal, the TIP tystem receives a command to retract the trave)ing probes. Upon full retraction, the isolation valves are then closed automatically. If a traveling probe were jammed in the tube run such that it could not be 5.2-11 Revision 4 - July 1984

I FNPS-F3AR retracted, instruments would supply this information to the operator, who would in turn investigate to determine if the shear valve should be operated. ) The two 18 in purge and vent line pipe entrances into the drywell have been provided with baffle plates to prevent debris from entering the lines during an accident. Any debris would threaten the ability to close the appli.able isolation valve. The Na makeup and vant isolation valves are used to relieve high drywell pressure during nonaccident conditions.

However, these s.

valves may be used after an accident provided the required power supplies are available and a low-low water level signal is not present. Esction 5.4.3 describes the Na makeup and vent valves used following an accident condition. I Lines, such as those of the RBCCWS which do not connect to the Reactor Primary System or open into the primary containment, are provided with at Icast one ac-powered valve on the effluent line and & check valve on the influent line. The Control Rod Hydraulic System is provided with three valves d.fch can be utilized for isolation purposes. The first is a bal: @,ck _ _ valve which comprises an internal portion of the control orive mechani=m. The other valves are norna11y closed hydraulic system control valves located in the Reactor Building. ) 5.2.3.5.3 ' Instrument Piping Connected to the Reactor Primary System Instrumentation piping connecting to the Reactor Primary SyLtem which leaves the primary containment is dead-ended at instruments loca t.ed i in the Reactor Building. These lines are providad with flow limiting orifices, manual isolation valves, and excess flow check valves. Inst rument sensing lines that originate within the reactor coolant pressure boundary and penetrate the primary containment are 1 in dia seismic Class I lines: 1/4 in dia orifices are installed in each of these lines inside the primary containment. This orifice size was selected to provide the same effective fluid cross sectional area as the excess flowcheck valves when fully open. A manually operated stop valve and excess flowcheck vaive are installed in each line insnediately outside, and as close as practicable to the primary containment consistent with the requirement for access to the stop valve. The combir.stion of orifice and excess flowcheck valve will reduce leakage to as low a value as practicable in the unlikely event of line failure. A failure of the excess flowcheck valve will result in a maximum leakage rate of 2 gal / min. A failure of the excess flowcheck valve body or the instrument line upstream of this valve would result in a maxistum leakage rate of 20 gal / min. In each of these instances the Icakage is well within the capability of the Reactor Coolant Makeup System. 5.2-12 Revision 4 - July 1984 1 I J

i l! PNPS-PSAR The amount of steam released to the Reactor Building from a j 20 gal / min leak would.not result in a failure of secondary containment. If the Reactor Building is not isolated, there would not be any. significant pressure rise due to the relatively high Reactor Building ventilation exhaust rates. If the Reactor Building is isolated, the operation of one standby gas treatment filter train I will prevent Reactor Building pressure from exceeding its design-value. An analysis of the potential offsite exposure that would result frorr a 20 gal / min leak into the Reactor Building has been performed. Such l a leak corresponds to an assumed failure of an instrument line outside the primary containment but upstream of the excess f1:)wcheck L valve. It was assumed in the. analysis that manual shutdown and ] depressurization would be initiated within 30 min. The. delay of I ~ i I i l l i 4 1 ) \\ 5.2-12a Revision 4 - July 1984 = = -

~ i ?NPS-FSAR 30 min is extremely conservative considering the numerous ways such a leak may be detected. The analysis assumed that steam frem the leak would be released to the environment through the normal ventilation path until the reactor had been depressurized. Based on these assumptions, the total dose at the site boundary for the duration of exposure was computed tc be 0.15 rem to the thyroid, which is substantially below the guidelines of 10CFR100. 1 Pressure retaining welds of instrument sensing lines that are part of the reactor coolant pressure boundary receive.ma gne tic particle or liquid penetrant examination of the Icst pass. Instrument line " bundles" are routed so as to minimize the potential for accidental damage. They are generally routed high in I compartments to ensure they are not stepped en or otherwise damaged. The lines are equipped with flow limiting orifices and excess flowcheck valves and are of the same size and schedule; therefore, the possibility of one line causing failure in another is extremely remote. The containment penetrations for these sensing lines are.shown on Figure 5.2-12. The 10 in drywell penetration sleeve contains six, equally spaced, 1 in, schedule 80 stainless steel instrument lines. J The manual isolation valves are 1 in stainless steel globe valves and are located as close to the penetration as practical, c o sistent with the need for access to the valve. The excess flowcheck valves close automatically on ' flow in excess of 2 gal / min. Neither ti.- manual isolation valves nor the excess flowcheck valves are equipped with position indicators. Regular monitoring of measured variables and comparison between redundant instruments provides operating personnel with sufficient information to identify malfunctioning or inoperative instruments and sensing lines. Operating and/or testing procedures will assure the operability of the safety related instrument lines and their associated orifices and excess flow check flows. An analysis was conducted to deternine the amount of Reactor Building ventilation that sould be required to prevent exceeding the design internal pressure of the Reactor Building for an instrument line blowdown through a 1/4 in orifice. The required vent;i.lation flow rate under these conditions is approximately 2,000 ft / min, which is 3 far below the available flow rate through either the normal Reactor Building Ventilation System or the Standby Gas Treatment System (SGTS). An instrument line failure will therefore not result in a le ss of integrity of the Secondary containment System (SCS)'. An estimate of the potential offsite exposure that would result from an instrument line failure has been calculated. The assumptions employed in this analysis were: 1. An instrument line failure occurs and results in an initial k blowdown of 2.2 h mass /see into the Reactor Building 5.?-13

f PNPS-FSAR 2. This blowdown continues undiminished and undetected for a period of 30 min 3. After a period of 30 min, the reactor is shut down, depressurized, and cooled down at a controlled 100'F/hr 4. The water-which flashas to steam is carried out of the Reactor Building by the normal ventilation system for the duration of the blowdown 5. The I-131 concentration in the blowdown is 6.1 x 10-2 microcurie /ml and the total iodine concentration is 1.G x 108 microcurie /ml 6. The atmospheric diffusion factor (X/0) for a ground level y release, 500 m distance to site boundary, and wake dilution factor of 3 is 5 x 10-4 sec/ma 7. The breathing rate is 3.47 x 10-4 m3/sec The above estiniates assume that corrective action would not begin for a period of 30 min. The detection of a sensing line break wou?.d be almost immediate by one or a combination of the means listed below. -- Proper corrective action would then be taken by the operating staff in accordance with station procedures such that the leak would be / \\ isolated or station shutdown and depre surization be initiated. It \\q,,/ is believed that it is not credible to assume no operator cetion ) would be taken in 30 min to terminate the consequences, and that the analysis based on a 30 min allowance for thes!: actions is very conservative. Sensing line break detection means are: 1. By a scram, annunciation, and possible instrument readouts and/or initiation of reactor safeguards systems if rupture i occurred on a Reactor Protection System instrument line ) 2. By annunciation of the control function, either high or low in the control room j 4 3. Operator comparing readings with several instruments monitoring the same process variable such as reactor level, jet pump flcu, and steam pressure 4. By increases in area temperature monitor readings and high temperature alarms in the Reactor

Building,

'and/or ventilation exhaust air ducts i 5. By a general increase in the area radiation monitor readings throughout the Reactor Building ~ 6. The leak should be audible either inside the Turbine j i Building or outside the Reactor Building to the operating staff members on a normal tour 5.2-14

• +

PHPS-ySAR 7. By det,ecting. the leak; as soon : as an access door to the Reactor Building is opened or approached - 4 Routine surveillance and the multiplicity of detection methods on the part of the operator as given. in items 1 through 7 above, represent an adequate means for detection of incipient or sudden failure of these small-diameter instrument lines and components. .5.2.3.6 -Venting and Vacuum Kelief System 1. General 1 j The. purpose of the vacuum relief valves is to equalize the pressrce between the drywell and suppression chamber and reactor building so that the. structural integrity of the containment is maintained. The vacuum relief system from the pressure suppression chanber to reactor building consists of two 100-percent vacuum relief breakers (2 ~ parallel sets of 2 valves in series). Operation of either system the l will maintain the pressure differential less than 2.5 psige ' ir - suternal design pressure. One valve may be out of service' for repairs for a period of 7 - days. If repairs cannot be completed within'7 days, the reactor coolant system is brought to a condition where vacuum relief is no longer required. The capacity of the 10 drywell vacuum relief valves are sized to limit the pressure ' differential between the suppression chamber and drywell during post-accident drywell coolant operations to the design limit of 2.5 psig. They are sized on the basis of the Bodega Bay l ,Ad pressure suppression system tests' '. The ASWi. Boiler and pressure Vessel Code, Section III,. Subsection B, for this vessel allows a 5 psig vacuum; therefore, with two vacuum relief valves secured an the closed position and eight operable valves, containment integrity is not impaired. Reactor operation is permissible if the bypass area between the primary containment drywell and suppression chamber does not exceed an allowable area. The allowable bypass area is based upon analysis considering primary system break

area, suppression chamber effectiveness, and containment design pressure.

Analyses show that the maximum allowable bypass area is 0.2 ft'. Reactor operation is not permitted if differential pressure decay rate is demonstrated to exceed 25 percent of allowable, thus in the event providing a margin of safety for the primary containment of a small break in the primary system. 2. Relief Valve Monitors The drywell to torus vacuum brsakers are installed to assure that the drywell pressure is at least equal to or greater than the pressure in the torus. In addition, when the vacuum breakers are in the closed

position, the drywell atmosphere (postulated steam) is directed through the suppression chamber downcomers during conditioM of drywell pressurization. To fulfill this engineered safety feature.

5.2-15 Revision 5 - July 19H

FNPS-FSAR 4 proper positioning and ' operation. of the vacuum breakers must be ensured. Therefore, each Pressure Suppression Chamber-Drywell Vacuum Breaker - is fitted with redundant pairs of position switches which l provide signals of disk position.to panel-mounted indicators and redundant' annunciators to alarm in the main control room if the disk is open more than the allowable limit. 5.2.3.7 Primary Containment Cooling and Ventilation System The Primary Containment (drywell) Cooling. System utilizes eight fan coil units distributed inside the drywell. See Figure 5.2-18. The . Frimary Containment Cooling. and Ventilation System design parameters are oiven on Table 5.2-2.. Each fan coil unit consists of two cooling coil-and two direct-connected motor-driven vaneaxial fans. Each cooling coil is connected to a cooling water supply and Ieturn piping - system inside the drywell. One or both cooling coils may be utilized l

• e-r i

1 l 1 1 5.2-15a Revision 2 - July 190 i 1 i i

l s PNPS-PSAR ) l l 1 for temperature control. Each unit recirculates the drywell atmosphere through the cooling coils to control the drywell space l r. temperature. Cooling water is supplied from the RBCCWS. Thervocouples are provided to monitor the performance of the drywell cooling system. They are installed in the air and water connections j of the drywell coolers as well as the air outiets of the reactor . vessel recirculate.on pump motors as shown on Figure 5.2-18. (The thermocouple, on water connections are also shown on Figure 10.5-1). , Temperature reaf. outs are provided on indicating panel C-2261A locate outside the drywell. J l Fan coil units circulate. cooled air around the recirculating pumps and motors, the control rod drive area, and the annular space between the reactor pressure vessel and the biological shacid. The personnel access and control rod drive remeval openings are sealed to ensure d pchative flow of cool air from the control rod drive area into the -e annular space between the reactor vessel and the biolccical shield, through pipe openings in the reactor vessel support locatrd primar21y at the upper level of ths control rod drivt space. Cooled air will also be circulated through the reactor vessel head; area-the space immediately below the ti. fueling seal plate, and the. -relief valve area. Each fan coil unit has provisions for installing dust filters. Filters are to be employed during drywell maintenance activities ard will be removed prior to normal station operation. Cooling water flow to esch coil is controlled independently by an electric motorized modulating v61ve positioned by a valve positior.er in the vontrol room. The cooling coil leaving car temperature car t.e adjusted by regulating the flow of cooling water. A cooling coil W failure can be detected by a flow device located in the cooling unit H condensate drain line, which is knnunctated in the mein control room. ) The standby coil can be put in tervice and the other isolated by their motorszed valves and a check valve in the return line. y Each fan is started from a local panel by using run-off-auto type I switches. One fan is started by switching to RPN and th$ other fan switch is placed in the AUTO position. If the normal operating fan q fails, a flow switch will sense a reduced pressure end automatically 1 start the standby fan, light an ember light <a t a local panel, and annunciate in the centrol room. Cooling vnit discharge air temperature is sensed by a temperature alement and indicated in the i control room. Upon scram, standby fans will be placed in service automatically to provide additional cc,oling. All fan coil units can be operated from the emergency pm,er supplies. The drytell purge ventilation supply system consists of. two full ( 4pacity fans to supply clean Reactor Building air t'> the drywell for purge and rentiletion purposes, during the reactor shutdown and refuelin( perieds for personnel access and occuponey. The purge Y exhaust air is normally discharged to the atmosphere through the i 5.2-16 Revision 4 - July 1984

) PNPG-FSAR Reactor Building exhaust vent. If necessary, SGTS is used for cleanup and the drywell air is exhausted through the main stack. The ventilation lines supplying the primary containment are provided with two fast acting, pneumatic, cylinder-operated butterfly valves in series for isolation purposes. These valves are normally closed during station operation. I a 5.2-16a Revision 4 - July 1984

1 1 l PMPS-FSAR 5 1 j Procedures for normal primary containment venting and purging are established such that gaseous effluent releases from the station { remain within the normal release limits. As noted above, purge or l vent exhausts can be directed to elevated release points through the I Reactor Building vent, or through the SGTS to the main stack. l Drywell and torus purging will normally be conducted to facilitate personnel access subsequent to periods of operation with the primary l containment ine rte d. Primary containment purge operations would normally release on the order of 1 million standard ft* of gas. 4 Drywell and torus venting is required during reactor startups in order to maintain normal operational primary containment pressure i control as heat loads increase drywell atmosphere temperatures. The I volume of gas released during venting operations is expected to be small with respect to purge volumes, f l Before purging or venting the containment, airborne contamination ] levels will be determined and estimates made of expected gaseous j activity releases. Selection of release routes and releare " rates j 5 will be made so as to assure compliance with the Tecnnical l Specifications. No special area controls or monitoring procedures i are imposed during primary containment purging or venting operations. 5.2.3.8 Primary Containment Atmospheric Control System l The Primary Containment Atmospheric Control System (PCACS) has been l provided in the design to introduce makeup air or nitrogen into the primary containment. J L.J The capability to operate the primary containment with an inert i atmosphere has been provided in the design in accordance with j previous licensing commitments. This system is capable of reducing ) and maintaining the oxygen content of the atmosphere and complies i with the requirements set forth by the American Gas Association. The l PCACS will be isolated from the primary containment in the event of I j an accident. Basically, the equipment in the PCACS performs two functions: (1) initial purging of the primary containment, and (2) providing an automatic supply of makeup gas. If the inerting system is used, the i purging equipment convarts liquid nitrogen into gaseous nitrogen. l Gaseous nitrogen can be introduced into the suppression chamber or l the drywell. The PCACS is also capable of automatically providing j makeup gas to the primary containmeYc. l 5.2.3.9 Drywell Temperature and Pressure Indication Drywell temperature and pressure are recorded in the main control rooc. These instruments can be utilized to eonitor the-essential drywell parameters tnat are used in the Station Safety Analysis in Section 14. a 1 1 5.2-17

l PHPG-PSAR j 5.2.'3.10 Pressure Suppression Pool Temperature and Levei Indication Pressure suppression pool' local: and bulk temperature is indicatd. ] recorded and alarmed in the main control room. Pressure suppression pool level is continuously. indicated in.the madn control rocm. These I instruments cary be 'utili2nd to monitor the essential pressure suppression pool' parameters that are assumed for initial values in the Station Safety Analysis in Section 24. 'j J 5.2.4 Safety Evaluation .5.2.4'.1 General q i i The primu y containment and its associated safeguard systems accomplish the following safety design bases: 1 l'. Accommodate the trats'i ent pressures and te.aperatures 1 associated with the postulated equapment failures within the containment (safety design basis 1) 2. Provide-at margin : for the effects of 2 metal water and other chemical reactions subsaquent to postulated accidents involving loss of coolant (safety design basis 2)- 3. Provide a high. integrity barrier against leakage of any fission products associ ted with these equipment failures t (safety design basis 3 and 5) \\ 4. Provide for long term core flooding (safety design basis 4) Y 5. Provide, for rapid actuation of the containment barrier (safety design basis 6) 6 '. Store water for the CSCS (safety design basis 7) 7. Maintain the containment paranatera during planned operation ~ to within those assumed in the Station Safety Analysis (safety design basis 8) These factors are, considered in the following evaluation of the integrated PCS. 5.2.4.2 Primary Containme,.t Onaracteristics Fo11cwing a Design Basis Accident In order to establish a design basis for the pressure suppression containment with regard to prassure and temperature rating and steam condensing' capability, tne :aaxircum rupture size of the Reactor ~ Primary System must be defined. For this dissign, an instantaneous. circumferential rupture with double ended flow of one recirculation line has been selected as a basis for determining the maxim'um grosa drywell pressure, and the condensing capability of the pressure 5.2-18 Revision 5 - July 19E5 f

T PNPS-FSAR suppression system. The selection of a failure of this size for the design basis is entirely arbitrary, since the cirewnferential failure of a recirculation pipe of this rnagnitude is considered to be of f i i. r ..s b' r. i 1 ) 1 5.2~18a Fwision 5 July 1955

( PNPS-FSAR exceedingly low probability. Nevertheless, for design purposes these failure condf.tions have been selected to establish the containment parameters. The design pressure is established on the basis of the Bodega Bay pressure suppression tests. The design pressure is primarily a l function of the postulated rupture area, the drywell to suppression l chamber vent ares and configuration, vent submergence below the water level in the suppression pool, and the final equilibrium pressure in l the pressure suppression chamber. 1 In establishing the containment design, circumferential pipe ruptures are assumed with sufficient distance separation to allow full potential f1Lv from each end of the pipe. Pipeline flow restrictions are not considered in establishing rupture flow rates. Since the assumed initial rupture rate and the accompanying reactor -y depressurization is so rapid, progressive failure of the piping is not a liniting factor in the containment design, The containment design parameters listed on Table 5.2-1 are concerned pr?marily with the effects on the primary c9ntainment caused by the i blowdown immediately following the postulated double ended rupture of the recirculation piping or equivalent failure. The parameters having the greatest effect or, drywell design pressure jL-) are the ratio of pipe break area to total vent area, the vent submergence below the water level in the suppression pool, initial system pressure, and the equilibrium pressure in the pressure suppression chamber before the postulated rupture. Sufficient water is p~rovided in the suppression pool to accommodate the initial energy which can be released into the drywell fro;n the ~ F etulated pipe failure..The suppression chamber is sized to contain .nis water, plus the water displaced from the Reactor Primary System together with the free air initially contained in the drywell. The primary containment response analysis to the design basis LOCA is presented in the Station Safety Analysis in Section 14. ] i It is concluded that safety design basis 1 is met. 5.2.4.3 Primary Containment C3pability The pressure of the PCS depends on both the system temperatures and the amount of noncondensible gases.

Thus, the capability of the system to store gases resulting from metal water reaction varies with the rate and extent of the reaction.

l Containment capabi14ty is defined in terms of the maximum percent of { fuel channels and fuel cladding material which can enter int 6 a metal 1 water reaction during a specified duration without the design pressure of the containment structure being exceeded. The analysis of the postulated LOCA discussed in the Station Safety Analysis in Section 14, shows that the operation of either of the two core spray I 5.2-19 7 a_,,. a_w,_,,__,_,_ __._____.__~_--a_-_+h-

PNPS-FSAR systems will maintain continuity of core cooling such that the ntent of the resultant metal water reaction is negligible.

However, to evaluate the containment system design capability various percentages

-l 1 of metal water reaction were assumed to take place over various durations of time. This anelysis presents an arbitrary method of measucing system capability without requiring prediction of the detailed events in e particular accident condition. The results are presented in section 14, Station Safety Analysis. It is concluded that safety design basis 2 is met. 5.2.4.4 Primary Containment Leakage knalysis The primary containment was tested to verify that the initial leakage rate was not in excess of the maximum allowable leak rate at the y calculated peak accident pressure. The maximum allowable leak rate was derived from the maximum allowable accident leak rate when corrected for the effects of containment environment under accident and test conditions. The resultant doses from the DBA for assumed leakage rates are i presented in Section 14. -- It is concluded that safety design basis 5 is met. p, 5.2.4.5 Containment Integrity Protection ) The PCS is designed for the loading considerations given in Sectien 12 and Appendix C. In addition, specid consideration is given to missile protection under the assumed accident conditions. The following summarizes the pertinent design considerations. ~ All large pipes which penetrate the containment are designed so that they have anchors or limit stops located outside of the containment to limit the movement of the pipe. These stops are designed to withstand the jet forces associated with the clean break of the pipe and thus maintain the integrity of the containment. Jet forces which may act on the containment are tah n as equal to reactor prescure acting directly on the containment over an area equal to the cross sectional area of the largest 2ocal pipe or nozr.le. The drywell is enclosed in reinforced concrete for shielding purposes and to provide aJditional re,sistance to deformation and buchling in l areau where the concrete backs up the steel shell. Whera concrete is i not available, such as at the vent openings, barriers are put across the6e openings for jet protection. d Based upon the conservative piping design utilJzing proven Engineering design proctice, the proper choice of piping ma t e r.icis. j the use of conservative quality control standards and procedures for piping fabrication and installation, and extensive studies of medes of pioe failure, it is concluded that pipes will not break in such a canaer as to bring about movement of the pipes cufficient to damage l the primarir containment vessel, i 5.2-20 L. -- _ _._______ a

PNPS-FSAR ) Although it has been concluded that with the application of conservative piping design and proven engineering prectices pipes will not break in such a manner es to bring about movement of pipes sufficient to damage the primary containment vessel, the design of the containment and piping systems does consider the possibility of missiles being generated frem the failure of flanged joints such es valve

bonnets, valve

stems, recirculation

pumps, and from instrumentation such as thermowells..

The most positive manner to achieve missile protection is through basic station arrangement such that, if failure should occur, the directisn of flight of the missile is away from the containment vessel. The arrangement of station components takes this possibt.lity !.nto account even though such m.issiles may not have enough energy to penetrate the containment. ir Spatial separation and utilization of the biological shield to the k maximum extent practical are the measures tcken to minimize the possibility of a single potential missile causing a loss of more than one redundant subsection of a vital safety system or a loss of more. than one functionally inderandent safety system. Ir> order to minimize post-accident containment leakage, the ~~ containment penetrations are designed to retain their integrity daring postulated accidents. . b It is concluded that safety desigr. basis 3 is met. 5.2.4.6 Containment Isolation j l One of the. basic purposes of the PCS is to provide a minimum of one protective barrier between the reactor core and the environmental surroundings subsequent to an accident involving failure of the piping components of the Reactor Primary System. To fulfill its role as a barrier, the primary containment is designed to remain intact before, during, and subsequent to any LOCA in a process system either inside or outside the primary containment. The process system and the primary containment are considered as separate systems, but where process lines penetrate ~the containment, the penetration design achieves the same integrity as the primary containment structure itself. The process line isolation valves are designed to achieve the containment function inside the process lines when required. l Since a rupture of a large line penetrating the containment and connecting to the Reactor Coolant System may be postulated to take place at the containment boundary, the isolation valve for.that line I is required to be located within the containment. This inboird valve in each line is required to be closed automatically on various indications of reactor coolant loss. A certain degree of additional reliability is added if a second valve, located outboa'rd of the containment and as close as practical to it, is included. This second valve a*so closes automatically on isolation signal. Both t valves shall receive the isolation (closure) signal even if normally l 1 1 5.2-21

PNPS-FSAR c'iosed during reactor operation. If a failure involver one valve, the second valve is available to function as the containment barrier. By physically separating the two valves there is lesc likelihood that a failure of one valve would cause a failure of the second. The two valves in series are provided with independent power sources. em The ability of the steam line penetration and the assosiaied steam line isolation valves to ft1 fill the containment objectives under several postulated break locations in the steam line is described below, and demonstrates the adequacy of the isolation valve design: 1. The failure occurs within the drywell upstream of the inner isolation valve l 3 Steam from the reactor is released into th'e drywell and tha resulting sequence is similar to that of a design basis LOCA except that the pressere transient is less cevere since the blowdown rate is slower. Both isolation valves close upon I receipt of the high drywell pressure signal or the signal indicating low water level in the reactor vesael. This action provides two barriers within the steam pipe passing through the penetration and prevents further flow of steam to the turbine. Thus when the two isolation valves close df[))1( f subsequent to this postulated

failure, the primary containment barrier is established, and the reactor is effectively isolatc6 from the external environment 2.

The failure occurs within the drywell and renders the inner isolation valve inoperable Again the reactor steam will blow down into the primary containment. The outer isolation valve will close upon receipt of the high drywell pressure or low water level signal, establishing the primary containment barrier 3. The failure occurs downstream of the inner isolation valve either within the drywell or within the guard pipe Both isolation valves will close upon receipt of either a high drywell pressure sign:- or a signal indicating low water level in the reactor vessel. The guard pipe is designed to accommodate such a failure without damage to the drywell penetration bellows, and the design of the pipeline supports protect its welded juncture to the drywell vessel. Thus the reactor vessel is isolated by the closure'Lf the inner isolation valve.and the primary containment barrier is established by closure of the outer isolation valve. It should be noted that this condition provides two ba,rriers between the reactor core and the external environment 4. The failure occurs outside the primary containment between 'the guard pipe and the outer isolation valve 5.2-22

i j PHPS-FSAR { ,1 The steam will blow directly into the pipe tunnel through the blowout panel and into the Turbine, Building until the j c' iso?ation valves are automatically closed. Closure of the j iraer isolation valve places a barrier between the reactor J core and the external environment. This barrier serves to l isolate che reactor and complete ' the containment integrity. i Closure of the outer isolatien valve in this ir.cidcnt sereres no useful purpose 5. The failure occurs outside the primary, containment and renders the outer isolation valve inoperative J J l The primary containment barrier and isolation of the reactor is achieved by closure of the inner isolation valve f 6. The failure occurs outsidc the primary co itainment between l i the outer isolation velve and the turbine 1-The steam will blow down directly into tho pipe tunnti or the Turbine Building until both isolation valves are automatically closed. This action isolates the reactor, establishes the primary containment barrier, and places two. barriers in series between the reactor core and the outside environment. l The exceptions to the arrangement of isolation valves described above ^~ y for lines conne:: ting directly to the containment or Reactor primary 'J n System are made only in cases where it leads to a less desirable i situation because of required operation or maintenance of the system { in which the valves are located. In the cases where, for iexample, 1 the tro isolae. ion valves rare located outside the containment, special { attention is given to assure that the piping to the isolation valves t has an integrity at least equal to the containment, f 1 Tne TIP system isolation valves are normally closed. When the TIP syrtem cable is inserted, the valve of the selected tube opens automatically and the chamber and cable are inserted. Insertion, calibration, and eetraction of the chamber and cable requires 4 ~. approximately 5 min. Retraction requires approximately 1 1/2 min. If closure of the valve is required during calibration, the isolation signal causes the cable to be retracted and the valve to close i i l automai;ically on completion of esble withdrawal. It is neither necessary nor desirable that every isolatico valve close simultaneously with a common isolation signal. For exaeple, if a process pipe were to rupture in the dryvell, it would be important to close all lines which are open to the dryy11, and some effluent process lines such as the main steam lines. However, under these conditions, it is essential that containment and core cooling systems f be operable. For this reason, specific signals are utilized for isolation of the various process and safeguard systems. Isolation valves must be closed before significant amounts of fission prodacts are released from the reactor core under DBA conditiona. 5.2-25 i 1 _____-_____-___-__Q

t 4 1 l PNFS-TSUi 1 f: 3 w,, ,1 Because the amount of' radioactive materials ir. the reactor coolant is small, - a sufficient limitation of fassion product release will be - accomplished iif the isolation valves are closed befort the coolant drops belo% the top of the core. it is' concluded that'safny-design basis 6 is met. 5.2.4.7 Containment flooding As is discussed in Section 12, the PCS is designed for the. conditions associated with flooding the containment. 1 It is concluded that safety design basis-4 is met, j 5.2.4.8 Pressure Suppression Pool Water Storage ~ d Based upon the-Station Safety Analysis presented in Section 14,. the quantity of water stored in the suppression pool is. sufficient to { ,Y condense the steam from o DBA and to provide water fe the CSCS. As discussed in-Section 12, the sup7r6ssion pool is considered in the loading conditions on the PCS. It is concluded that safety design basis 7 is met. 5.2.4s9 Limit'ations During Planned Operations ~ ~ As is discussed in Sections 5.2.3.6, 5.2.3.7, and 5.2.3.8, the PCS is ( I (D) designed to be kept within the limits of. parameters assumed in the Static.n Sasety Analysis presented in Section 14 during planned operations.- I It is concluded that safety design basis 8 is met. ~ '5.2.4.10 Primary Containment Stec.m Quenching i i The suppression ' chamber, or torus, is' designed to contain a pool of water in. order to supprass the prsssure during a postulated LOCA by j condensing the steam released from the Reactor Primary System. The reactoe system energy released by relief valve operation during r L operating transients also is released into the suppression pool As a result of cencerns ~ regarding potential instability cf steem condensation in "a het ' suppression pool, the United States Nuclear Regulatory Commission (NRC) has imposed pool temperature limits for rjlant t'cansients . involving safety / relief valvs (SRV) operation 4 (Reference 1). The limits which ensure smooth steam condensation for dischargk through quenchers are: J

1.. For all plant transients involving SRV operatAun during which the steam flux through f.he quencher perforations exceeds 94 lbm/ft*

-sec, the suppression pool local temperature shall not exceed 200*F. J 5.2-24 Revision 5 - July 1985 .A. I

jffs w: L ', ep ~ FNp5-FSAR e

s.,

.s C 2. ForL all plant transients'during which the ste6m flux through the cc

quencher. perforations;. is less

- than' 142 lbm/ft*-sec. the-suppression pool local temperature should be at. least 320*F subcooled. 'This corresponds to: a local temperature limit of '.201.4*F for pNPS. 3. For plant transients. involving SRV operation during which the steam flux through the quencher perforations exceeds 42 lbmtft* g -see but! is less than 94 lbm/ft -sec, the suppression pool local-temperature can beLestablished by linearly interpolating the local temperatures established under items (1)'and (2) s.bove. These' limits.are depicted in Figure 5.2~19. t An.ahalycis :vas dona 1 Reference 2)'to show that pNPS complies with. the NRO cri+.eria.- l Y. -) ~ Seven.. transient ; events have been identified, one of which is expect,sd } to result in the maximum long-term suppression pool temperature. The seven events'are as;follows: Stuck-open SRV during power operation with one RHR loop 1A. 1 available. 1B. Stuck-open SRV during power operation.. assuming reactor i isolation due to MSIV closure. JA. Isolation / scram and manual depressurization with bne RHR loop available.

23.. Isolation / scram and manual depressurization wit'h the failure 1

of an SRV to reclose (SORV). 2C. Isolation / scram and egnua"; depressurization.with two RHR . loops .available. This case demonstrates the p6el temperature responses when an isolttian/ scram eernt oe. cure under normal power operation (i.e.. when all Systeas' are operati.ng in normal made). 3A. Small-break accident (SBA) with manual depressbrication: accident mode with one RHR loop available. 3B. Small-break-accident (SBA) uith manual depressurization and failure of the shutdown cooling systtm. The analysis indicated that the maximum temperature occurs during Cuse 2A. The seximum local pool temperature for this cese, is 199'T (ieference 4) which is less than the 201.t*F limit applicsble for low steam flux conditions. 5.2-04a Revision 5 - July 1985 w

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l y PNPS-FSAR TABLE 5.2-1 PRIMARY CONTAINENT SYSTEM PRINCIPAL DESIGN PARAMETERS AND CHARACTERISTICS i Pressure suppression chamber: Internal design pressure +56 psig External design pressure +2 psig Drywells i Internal design pressure +56 psig External design pressure +2 psig l (approx) 147,000 ft8 Drywell free volume i. Pressure suppression chamber free volume (approx) 120,000 ft3 Pressure suppression pool water volume, maximum .(approx) 94,000 fts - Pressure suppression pool water volume, minimum .(approx) 84,000 ft8 Submergence of vent pipe btRow pressure (approx) 3.00 to 3.25 ft suppression pool surface. Design temperature of dryvell 281*F Design temperature of pressure suppression chamber...... 281*F ~ Downcomer vent pressure loss factor . 6.21 Break area / total vent area. . 0.0194 Drywell free volume / pressure suppression chamber free volume.. 1.34 Primary system volume / pressure suppression pool volume. 0.268 Drywell free volume / primary system volume 7.4 Calculated maximum pressure during blowdown: 45 psig Drywell 27 psig Pressure suppression chamber. Initial pressure suppression chamber temperature rise . ~.. 35'F 1 of 1 Fevision 2 - July 1983

PNPS-FSAR t TABLE 5.2-2 DRYWELL ATMOSPHERE COOLING DATA SHEET Location Average Maximum

• r General 135'F 148'F Recirculation Pump Motor Area 128'F Entering Air Temperature to Cooliry Units 135*F 148'F Leaving Air Temperature from Cooling Units 85*F 95'F 5

Cooling Mater Supply Temperature 75'F 85'T Cooling Water Return . Temperature 90'F 100*F Drydell Heat Gain 2.4 X 108 Btu /hr 3.4 X 108 Btu /hr Total Cooling Unit Capacity 3.6 X 10s Btu /hr 5.6 X los Btu /hr g ) Total Cooling Unit Fan capacity 72,000 ft / min 110,000 fts/ min 3 Total Fan Brake hp 54.8 67.8 Drywell Temperature 10 hr after shutdown 105'F 105'F NOTE:

• As a result of higher cooling water supply temperature and extra heat load from scram of the control rod drives.

e 9 1 of 1

PMPS-TSAR therefore be concluded that the resultant radiological exposures for the above pipe failures will at the maximum be based on only that activity contained in the primafy coolant, which is discharged to the secondary containment. To provide an upper limit to the radiological exposures, the assumptions have been made that: 1. All of the primary coolant which contained. activity is l eventually discharged to the secondary containment l 2. Considering the thermodynamics of the coolant discharged, a maximum of 1/3 of the coolant is flashed to steam resulting in the release to the secondary containment of 1/3 of the coolant activity 3. Consideration of the condensing and plateout surfaces that i the released steam will have to come in contact with prior I to being released from the top of the Reactor Buf1 ding results in a minimum reduction factor of 3 for the rel, eased iodine activity 4. The activity is released from the top of the Reactor Building under those meteorological conditions, which ; maximize the offsite exposures The activity contained in the reactor coolant is consistent 5. with an offgas emission rate of 10' microcuries/see Based on the above considerations, the resultant site boundary thyroid dose is 0.08 rem while the LP thyroid dose is 0.002 rem. If the conservative assumption is made that downwash of the released effluent occurs and that the coolant activity is at a level consistent with the technical specification offgas activity (i.e., ~ 0.9 ci/see), the resultant site boundary thyroid dose is 15 rem and the LPZ thyroid dose is 0.6 rem, both of which are well below the 300 rem guideline set forth in 10CTR100. 5.2.9 References i 1. Bodega Bay Preliminary Hazards

Report, Appendix I,

i Docket 50-205, December 28, 1962, 2. General Electric

Company, "PNPS Unit 1 Suppression Pool Temperature Response," NEDC-22059-P, March 1982.

l l 3. J. M. Caroll, BECo Letters to NRC, May 15, 1973. i i 4. General Electric Comapny, " Pilgrim Suppression Pool Temperature i Analysis," Letters from R. Thibault to G. McHugh, Decembe'r 1982. l 5.2-37 Revision 5 - July 1995 .__._______.___________._m__-_.

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PHPS-FSRR 1 l l 6.

Jacobs, I.M., Quide-Describes methods used to lines for Determining establish allowable repair Safe Test Intervals times and Repair Times for Engineered Safeguards.

General Electric Co., Atomic Power Equipment Department, APED-5736, April 1969 l Each detailed requirement in the following analysis is referenced, if l possible, to the most significant station condition originating a l need for the requirements by identifying a matrix block on one of the Matrix 3 sheets of Table G.5-3. The matrix block referenced is given in parentheses beneath the detailed requirements in the " minimum required for action" section. 'T The matrix block references identify the BWR operating state, the event number, and the system number. For example, F39-82, identifies BWR operation state F (Matrix 3), event (row) No. 39, and system (column) No. 82. Minimum Required for Action l 1. Vacuum Relief System (C39-109) (E39-109) (D39-109) (F39-109) w 2. Primary Containment (C39-82) (E39-82) (D39-82) (F39-82) l 3. Drywell Pressure and Temperature Indicators ~ (C2-103) (E2-103) (D3-103) (F4-103) 4. Pressure Suppression Pool Water Level and Temperature Indicators (A35-104) (D3-104) (F4-104) (B35-104) (D39-104) (F39-104) (C2-104) (E2-104) (C39-104) (E39-104) 5. Pressure Suppression Pool Water Storage (A35-83) (D39-83) (B35-83) (E39-83) (C39-83) (F39-83) Proposed Limiting Conditions for Initial Plant Operation' The limiting conditions for operation are set forth in the Technical Specifications referenced in Appendix B. t 5.2-29 l 1

PNPS-FSAR With the nuclear system pressurized in states C, D, E, and F, the drywell and pressure suppression chamber shall be closed and all of the following conditions shall be satisfied: 1. All nonautomatic primary containment isolation valves which are not required to be open for station operation shall be closed 2. At least one door in the airlock shall be closed and sealed 3. The operational requirements governing isolation valves for the various categories of penetrations apply as listed in Section 7.3 4. All blind flanges and manways shall be closed 5. Pressure s0ppression pool water level and temperature y indications and drywell temperature and pressure indications shall be available The suppression pool water volume and temperature shall be maintained within the following limits: 1. Minimum water volume 84,000 ft3 (all states) corresponding to a water level of 2 ft 9 in below torus centerline i l 2. Maximum water volume 94,000 f t3 (states C, D, E, and F only) / i i corresponding to a vater level of I ft 9 in below torus centerline 3. Maximum pool temperature (1) during normal operation 80*F, ) (2) during accidents and transients 170'F immediately following depressurization; depressurization shall begin before t'ie pressure suppression pool temperature reaches 130*F 4. Minimum pool temperature shall not be less than 40'F In operating states C, D, E, and F, nine of the ten drywell suppression chamber vacuum breakers shall be operable. Both suppression pool Reactor Building vacuum breakers shall be operable except that one of the vacuum breakers may be inoperable for 1 month when primary containment integrity is being maintained. The setpoint on the differential pressure instrumentation which actuates the block valves that place the Reactor Building to primary, containment vacuum breakers in service shall not exceed 0.5 psi. The primary containment leakage rate as determined by test shali not exceed 1.6 percent of the contained volume per day at 56 psig, or 2.0 percent of the contained volume per day at peak accident conditions. 5.2-30 Revision 2 - July 1983 \\

PNPS-FSAR After complete installation of all penetrations in the drywell and suppression chamber, the vessel was pressurized te the, calculated peak accident pressure, and measurements taken to verify that the integrated leakage rate from the vessel did not exceed the maximum allowable leak rate. A second test was run at reduced pressure to establish a relationship between leakage rate and containment pressure. The necessary instrumentation is installed in the station to provide the data required to calculate and verify the leakage rate. Provisions are made so that integrated, containment leakage rate tests may be periodically performed during periods of reactor shutdown, in compliance with 10CFR50, Appendix J, Primar3 Containment Leakage Testing for Water Cooled Power Reactors. I 5.2.5.2 Penetrations The design permits the testing of penetrations which have resilient seals or expansion bellows without pressurizing the entire containment system. Leak detection may then be accomplished either by the use of soap suds, pressure decay techniques, or other _ acceptable methods. Pipe penetrations which must accommodate thermal movement are provided with two ply bellows expansion joints. These two ply bellows are provided with test taps so that the space between the plies can be pressurized to the calculated peak accident pressure to permit testing of the individual penetrations for leakage. Electrical penetrations are also separately testable. The test taps ~ are located so that the tests of the electrical penetrations can be conducted without entering or pressurizing the drywell or suppression chamber. All containment closures which are fitted with resilient seals or gaskets are separately testable to verify leaktightness. The covers on flanged closures, such as the equipment access hatches, the drywell head, access manholes, and personnel air lock doors, are provided with double seals, and with a test tap which allows pressurization of the space between the seals without pressurizing the entire containment system. 5.2.5.3 Isolation Valves The test capabilities which are incorporated in the PCS to permit leak detection testing of containment isolation valves are separated into two categories. ~ The first category consists of those pipelines which open into the containment atmosphere and do not terminate in closed loops' outside the containment, and contain two isolation valves in series. Test taps are provided between the two valves which pe rmit leakage i monitoring of the first valve when the containment is pressurized. 5.2-27

( I l l PNPS-FSAR 9 l The test tap can also be used to pressurize between the two valves to permit leakage testing of both valves simultaneously. The second category consists of those pipelines which connect to the Reactor System and contain two isolation valves in series. A leakoff line is'provided between the two valves, and a drain line is provided downstream of the outboard valve. This arrangement permits monitoring of leakage on the inboard and outboard valves during Reactor System hydrostatic tests, which can be conducted at pressures up to the reactor system operating pressure of 1,000 psig. Isolation valve closing time is determined during the functional performance test performed prior to reactor startup. l 5.2.6 Proposed Nuclear Safety Requirements for Initial Plant operation -v The entries in this section represent the proposed nuclear safety requirements for the PCS for each BWR operating state which represents an extension of the stationwide BWR systems analysis of Appendix G. The following referenced portions of the safety analysis report provide information justifying the entries in this section: Reference Information Provided ,-m 1. Preceding portions of Description of PCS '_s s Section 5.2 2. Section 7.2, Description of PCICS Primary Contain-ment and Reactor Vessel Isolation Control System 3. Station Safety Anal-Analyses verifying primary ysis, Section 14 containment responses and radiological effects of pos-tulated accidents 4. Station Nuclear Identification of condi-Safety operational tions and events for which Analysis, Appendix G PCS is required 5. Bodega Bay Prelim-Pressure suppression test inary Hazards Sum-information mary Report, Appen-dix 1, Docket 50-205 S 5.2-28

Ji PMPS-FSAR 5.2.4.11 Steam Bypass o -{ Following a reactor coolant pipe break inside the containment, the ] potential exists for the air-steam mixture within the drywell to pass through various leakage paths into the suppression chamber, thereby causing pressurization of - the suppression chamber. This increased back pressure in the s'uppression chamber might lead to an increase in pressurization of the drywell, and possible overpressurization of; the containment beyond the design limits. r The bypass area is expressed in terms of.A/sk, where A is the total bypass (leakage) area and k is the pressure less coefficient. The maximum allowable leakage area that could exist between the drywell and the suppression chamber is a function of the area of the break as well as the duration of pressurization. The former depends on the P between the drywell and suppression pool, and the latter relates to the time delay until containment sprays are initiated. 1 i.., In order to assess this relationship, an analysis was performed with various steam break sizes. For large breaks the AP is high, but has -i a short duration. The maximum AP results from the DBA. Primary { system-breaks greater than approximately 0.3 ft* will result in rapid depressurization of the primary system. Figure '5.2-22 r shows the allowable bypass capacity ( A/s k) as a function of primary system break area. The allowable A/s'k is determined on the basis of.the allowable steam ,;c mass that can be bypassed - without exceeding the containment ~ design pressure of 62 psig. For the Pilgrim Nuclear Power Station the maximum allowable bypass capacity is an A/N k =0.13 f t *. Typically, the geometric loss factor would be 3 or greater. Thus,'the actual 6 O 1 5.2-25 Revision 5 - July 19"5 ) i l ~

PNPS-FSAR [ allowable bypass area would be approximately 0.2 ft*. This is equivalent to a 6 in orifice. I When calculating the allowable leakage capacities shown on Figure 5.2-22, the following sequence of events is assumed. Immediately af ter a break in the primary system, a rapid rise in containment pressure would occur as the noncondensible gases in the drywell are transferred to the suppression chamber. For the allowable leakage calculations, no operator action is assumed until the suppression chamber pressure reaches 35 psig. Further, a 10 min delay is assumed before any action is taken to terminate the transient. In addition to the 10 min operator delay, a 5 min delay is assumed for corrective action to become effective. The following assumptions were made in calculating the allowable leakage espacities: 1. Flow through the postulated leakage is pure steam. This is -r a conservative assumption as the amout t of steam rel' eased into the suppression pool is maximized 2. There is no condensing of the leakage flow en ' either the suppression pool surface or the torus and vent system r structure. This assumption results in a conservative peak pressure calculation Station emergency procedures ensure that operator corrective action ( appropriate to the postulated events is taken. If the low-low water 1evel point has not been reached, the operators can depressurize the reactor vessel through the main steam lines to the main condenser or alternately, utilize the relief valves to rapidly reduce reactor pressure. Existing emergency procedures require the initiation of the pressure suppression pool spray mode of the RHRS after verification that the reactor vessel water level is adequate. Further, the procedures require the initiation of the drywell spray mode of the RHKS if the drywell pressure rises to 10 psig. 5.2.5 Inspection and Testing The following discussion details the surveillance and testing that will be conducted on the various systems or components of the primary containment during construction or station operation. 5.2.5.1 Primary Containment Integrity and Leaktightness Fabrication procedures, nondestructive testing, and sample coupon tests were made in accordance with the ASME Code of Boilers and Pressure Vessels, Section III, Subsection B. The integrity of the Primary Containment System was verified during cons t ructi'on. The verification included a pneumatic test of the drywell and suppression chamber at 1.25 times their design pressure in accordance with code j requirements. 5.2-26 1

PNPS-FSAR drywell areas requiring protection are shown on Figure 5.2-23, and are those areas in the spherical section where the single postulated i large pipe veld failure could result in pipe movement to the extent that the ruptured pipe could contact the interior of the drywell shell with sufficient energy to perforate the drywell. The protection system consists generally of steel members attached to a reinforcing plate. The protection system is arranged to receive a postulated rupture pipe, absorb a portion of the impact energy, and distribute the impact load over an area of the drywell shell such that the combined energy absorbtion capacity of the protection system and the drywell shell is greater than the impact energy of the ruptured pipe. The protection system and the drywell shell will deform through the 2 in air gap between the drywell and the concrete shield without causing breaching of the drywell. Details of the protection system are shown on Figure 5.2-23. Areas of the spherical section of the drywell shell requiring protection have been determined by plotting the potential area at which each ruptured pipe end could contact the drywell when the pipe is rotated around various possible plastic hinge points. The force: causing pipe movement and deformation around a plastic hinge is the . jet reaction resulting from blowdown from the reactor system. ~ The impact energy of a ruptured pipe has been determined as a 3 function of the jet reaction force, pipe plastic bending moment, and the configuration of the pipe with respect to the drywell shell. The energy required for perforation of the drywell shell has been de te rmined from an empirical relation developed from a series of experiments using steel projectiles. The protection system component size and placement is based upon the requirement to distribute the pipe impact energy over a sufficient area of the drywell shell, such that the combined energy absorbing l capability of the protection system and drywell shell is greater than l the impact energy of the ruptured pipe. The steel beams are arranged to minimize the possibility of a ruptured pipe end from resulting in I & localized load bearing directly on the drywell shell. The beams are attached to a steel plate located between the beams and the drywell shell. This plate results in increasing the energy absorbing capability of the drywell by increasing the impact area, and i increasing the effective thickness of the drywell shell. The plate also serves as a means of restraining the beams against potential jet impingement loads and the component of pipe impact loads tangential to the drywell shell. The protection system is attached to the drywell shell at the weld pads with additional support from the floor structures as required. The protection system supports are designed to withstand the loads from the Safe Shutdown Earthquake. 1 l The protection system components have been selected and located such ( that maximum protection is provided for the drywell shell against 5.2-35

f PNPS-FSAR postulated pipe ruptures with minimum interference to required access for inservice inspection. Pipe ruptures within the cylindrical section of the drywell have been considered and no protection is required because: i 1. Pipe movement distances to contact the drywell are insufficient to obtain an impact energy exceeding the energy { required to perforate the 1 1/4 in shell thickness 2. The close proximity of the drywell shell to the piping systems is such that pipe rotation around 6 plastic hinge is insufficient to result in the ruptured end becoming a localized load on the drywell The analysis and basis for design of the protection system is ,I conservative because: 1. Jet reaction forces have not been reduced due to the throttling effect of partial pipe closure at the plastic hinge point 2.. Pipe impact energies have not been reduced by the energy absorbed by pipe deformation at the point of contact between the protection system and the drywell shell ,(~' \\&- 3. Impact energy required to perforate the drywell shell is based on test data using tool steel projectiles, and is therefore lower than the energies required for perforation with typical pipe materials ~ 5.2.8.3 Design Basis Line Break The design basis steam line break accident (SLBA) is described in detail in Section 14. This accident is assumed to result in a complete guillotine break of the main steam line, resulting in a 1.74 f ta break area. All other breaks in piping attached to the vessel above the core result in peak clad temperatures which are lower than those resulting from the SLBA. Since there are no perforations for the SLBA, there will be none for smaller steam line breaks. While the SLBA evaluated in Section 14 considers isolation of the reactor vessel, an analy, sis of an SLBA inside the primary containment (i.e., no isolation) is described in Section 6.5.3.2. The results of this analysis show that the core will remain covered throughout the transient and the resultant peak clad temperature will be less than normal operating temperatures, which are well below the temperature where clad perforation could occur. As in the case. of the design basis SLBA, all other smaller steam lines which could fail in such a manner that isolation is not achieved would also not result in clad perforations. Consideration of those liquid breaks chich could conceivably result in containment breaching as a result of pipe whip has also resulted in the conclusion that no fuel perforation: ) will occur. In particular, for the feedwater line break (approximately 0.5 ft ) incore is also not uncovered. It can 2 5.2-36

FNPS-FSAR l rates in the downcomers, and to provide protection against negative pressure conditions in the containment vessel. Allowing one valve to be inoperative reduces the total vacuum relief area by only 10 percent. Since the valves are totally enclosed within the containment, possible leakage through them does not affect the containment system leakage. The Suppression Pool Reactor Building Vacuum Relief System assures that the primary containment is not operated at a significant negative pressure relative to its surroundings. The 0.5 psi differential pressure setting was chosen on the basis of Relief System pressure drop, valve opening times, and peak mass flow to limit the external pressure on the suppression chamber to less than its design value of 2.0 psig. The Vacuum Relief System is O redundant system and full relief capacity is available through either valve. If one vacuum breaker or its block valve becomes inoperable, there is no immediate threat to primary containment integrity,

thus, i

reactor operation may continue while repairs are being made, provided the repair procedure does not violate primary containment integrity. Possible leakage of these valves is included in the containment system integrated leakage rate tests performed periodically. 5.2.7 Current Operational Nuclear Safety Requirements The current limiting conditions for operation, surveillance i requirements, and their bases tre contained in the Techn2 cal Specifications referenced in Appendix B. 5.2.8 Pipe Break Transient Analysis i 5.2.8.1 Pipe Mechanical Failure and Safety Design ~ The Pilgrim Nuclear Power Station primary containment satisfies safety design basis 1 by its capability "to accommodate the transient pressures and temperatures associated with the postulated equ' ment failures within the containment." The intent of safety design basis 1 is to provide a basis for determining the primary containment internal design pressure and associated temperature, and that basis is that the primary containment must remain functional after accommodating the largest mass flow and energy release associated with the design basis LOCA. See Section 5.2.4.1. Section 5.2.4.2 discusses the selection of the failure conditions that establish containment design parameters. The capability to satisfy safety design basis 1 is demonstrated in the primary containment response analysis to the design basis LOCA present in the Station Safety Analysis in Section 14. It is not the intent, nor has it ever been assumed that it would be the intent, that safety design basis 1 be used as a basis for evaluating the abilities of the primary containment to accommodate the mechanical forces and energies that might be associated with the movement of unrestrained pipe during a postulated LOCA. 5.2-33

PNfS-FSAR Section 5.2.4.5 discusses containment integrity protection within the scope and the intent of safety design basis 3 and defines the pertinent loading considerations that have been evaluated in order to meet safety design basis 3. In order to minimize the probability of an instantaneous failure in the Reactor Primary System piping, design provisions were made to minimize or identify conditions that could lead to such a failure. As discussed in Appendices J.2.4 and F.2.6, the Reactor Primary System is designed to meet the intent of Criterion 35 of the Proposed A5C General Design Criteria, thus reducing even further the extremely low probability of an instantaneous piping failure due to brittle fracture. In addition, a Nuclear System Leak Detection System, as described in .i Section 4.10, is provided to identify primary system leakage rates well below those leakage rates which correspond to the critical size for rapid crack propagation. The capability of this system to identify these leakage rates will provide significant protection against an instantaneous primary system piping failure due to crack propagation by allowing station personnel sufficient time to take appropriate corrective measures. Supplementary protection will be provided tur the comprehensive inservice inspection program discussed s in Appendix K. In conclusion, the design fabrication, testing, and inspection of the Reactor Primary System has emphasized the elimination of potential causes for instantaneous piping failures, and thus obviates the need to design the PCS to withstand the mechanical effects of the failed pipe. Therefore, the protection of the primary containment from the mechanical effects of an unrestrained failed pipe is not a safety design basis for Pilgrim Nuclear Power Station. 5.2.8.2 Pipe Protection System As a result of an investigation, selected areas of the interior of the drywell shell were protected to reduce the possibility of breaching of the primary containment by postulated failure of a large, unrestrained pipe in the primary pressure boundary. All pipe penetrations through the drywell have been designed to withstand the forces and moments resulting from a pipe rupture inside the drywell. Main steam and feedwater piping have restraints outside the drywell for protection of the penetration assemblies and outboard isolation valves. The piping systems considered for postulated failure and having the potential to breach the containment are those which are located within the spherical section of the drywell and normally pressurized to reactor pressure (main steam, HPCI steam supply, feedwater, RHR). These large pipes are postulated to fail at circumferential butt h welds with the jet reaction force acting normal to the rupture

surface, and resulting in pipe rotation around a plastic hinge. The 5.2-34

I j \\ =. - PNPS-FSAR l l 1 l During planned operations. the mean dryweli temperature shall not J exceed 150'T and internal drywell pressure shall not exceed 2.5 psig. l Proposed Surveillance Requirements for Initial, Plant Operation Integrated leakage rate tests shall be performed on the primary containment, and the allowable test leakage rate shall not exceed 1.2 percent per 24 hr at a pressure of 28 psig. Measured reductions I in leakage obtained by repairs made immediately prior to or during the integrated leakage rate test shall be added to the test results. The allowable test leakage rate is used to determine the containment leakage retest schedule. Periodic integrated leakage rate tests at initial pressure of approximately 28 psig shall be performed at the following minimum frequency: l 1. During the first refueling outage l 2. Within 2 yr of test 1 above 3. Within 4 yr of test 2 above 1 4. Within 4 yr from the date of the previous test 5. In the event the primary containment leakage rate of any one test exceeds the allowable test leakage rate, the condition shall be corrected, a retest (local or integrated) made, and the testing frequency shall revert to the following schedule: a. Within 1 yr of retest [ b. Within 2 yr of test a above c. Within 4 yr of test b above d. Within every 4 yr from the date of the previous test 6. The preceding test schedules may be extended by t.ot more than 8 months. The ellowable operational leakage rete shall not exceed 0.9 percent per 24 hr at 28 psig. Following each test, if the measured leakage rate exceeds the allowable operatier.a1 leakage rate the conditions shall be corrected and a retest (local or integrated) made. Operation of the reactor under pressuri::ed conditions in operating states C, D, E, and F shall not be resumed until the leakage rate is less than the allowable operational leakage rate. 4 Primary containment testable penetrations and testable isolation valves shall be tested at a pressure of 56 psig each refueling outage, except bolted double gasketed seals shall be tested whenever 5.2-31 Revision 5 - July 19E5 { 1

PNPS-TSAR .'a j the seal is closed after being opened, and at least at each refueling outage. The setpoint of the pressure instruments which actuate the Reactor Building to primary containment vacuun breakers shall be checked once every 3 months. Operability of the Reactor Building to primary containment vacuum breakers shall be checked during each refueling outage. Operability of the suppression chamber to drywell vacuum relief valves shall be checked during each refueling outage. Operability of the drywell pressure and temperature indicators and suppression pool water level and temperature indicators shall be checked during each refueling outage. Requirements are placed on the operating status of systems essential to containment to assure their availability to control the release of any radioactive material from irradiated fuel in the event-of an accident condition. The PCS provides a barrier against uncontrolled release of fission products to the environs in the event of a break in the Reactor Coolant Systems. Whenever the reactor is in states 0, D, E, and F (with nuclear system pressurized), failure of the Reactor Coolant System could cause rapid expulsion of the coolant from the reactor, with an associated pressure rise in the primary containment. Primary containment is required, therefore, to limit the release of e fission products to the station environs so that offsite doses would j be well below the values specified in 10CTR100. v l The calculated radiological doses given in Section 14 were based on an assumed leakage rate of 0.5 percent. Increasing the assumed leakage rate at 56 psig to 2.0 percent, as indicated in the limiting conditions for operation, would increase those doses by approximately ~ a factor of four, still leaving a substantial margin between the calculated dose and the 10CTR100 regulation. The suppression pool water volume provides the heat sink for the Reactor Coolant System energy released following the LOCA. In states A and B the suppression pool water is available as a source of makeup water to replace possible leakage from the reactor vessel and primary system. The maximum water volume limit allows for an operating range without significantly affecting the accident analyses with respect to f ree air volume in the suppression pool. The maximum pool temperature of 130*F is permissible since a complete accident blowdown can be accommodated with minimum water volume without exceeding the temperature limit of 170*F immediately af ter blowdown. Th'e minimum water temperature of 40*F is specified to assure that the water is always in the liquid state. The Drywell Suppression Pool Vacuum Breaker System is required to prevent water oscillation in the downcomers due to low steam flow 5.2-32

o PNPS-FSAR TABLE 5.2-3 (Cont) 1 i ] Penetration Number Sleeve Number Required Size (in) Description X-102 A&B 2 12 Indication and Control ,X-103 A 1 12 Indication and Control X-103 B 1 12 Computer and Thermocouple X-104 A-J 9 12 CRD Rod Position Ind., TIP System X-105 A&B 2 12 480 V Power X-105 C 1 12 Containment High Range Radiation Monitor X-106 A 1 12 Recirc. Line Sample (X-106 A-A), Drywell Pressure, (B), Core Flow (C) Spare (D,E,F) y

. 105 B 1

12 Indication & Control X-200 A&B 2 48 ID Access Hatch X-201 A-H 8 Vent Lines X-202 A-B 2 12 Indication Power-Lights, Dew Point, Temperature X-203 A-K 10 18 ID Vacuum Breakers V-205 1 20 Torus Purge Inlet X-206 A-D 4 1 Torus Water Level Indication X-207 A-H 8 1 Vent Line Drain X-208 A-D 4 12 Relief Valve Discharge X-209 A,C 2 1 Torus Water Temperature X-209 B,D 2 1 Spare X-210 A&B 2 16 Containment Cooling and Core Spray Test Line X-211 A&B 2 6 Containment Cooling to Spray Header X-212 1 18 Spare ~ X-213 A&B 2 8 Construction Drain X-214 1 4 Spare X-215 1 4 Spare X-216 1 2 Spare X-217 1 2 Spare X-218 1 10 Spare X-219 1 10 Spare X-220 1 6 RCIC Pump Suction X-221 1 16 HPCI Pump Suction X-222 A-D 4 18 RHR Pump Suction X-223 1 24 HPCI Turbine Exhaust X-224 1 2 HPCI Turbine Exhaust Drain X-225 1 8 RCIC Turbine Exhaust X-226 1 2 RCIC Vacuum Pump Discharge X-227 1 20 Torus Purge Exhaust & Vacuum Relief l-X-228 A 1 1 Torus Pressure X-228 B 1 1 Reference Vessel Pressure 3 of 4

.,0 PNPS-FSAR TABLE 5.2-3 (Cont) Penetration Number Sleeve Number Required Size (in) Description X-228 C,H,K 3 1 0xygen Analyzer Suction & Return X-228 E 1 1 Vacuum Breaker Air Supply X-228 D,F,G,J 4 1 Spare X-229 A&B 2 18 Core Spray Pump Suction X-230 1 8 Spare X-240 A&B 2 1 Torus Water Level Monitoring System (DPT-1001-604B) X-241 A&B 2 1 Torus Water Level Monitoring "y System (DFT-1001-604A) \\ 1 4 of 4 Revision 2 - July 1983

] PNPS-FSAR i i 5.3 SECONDARY CONTAINMENT SYSTEM 5.3.1 Safety Objective The safety objective of the Secondary Containment System (SCS), in conjunction with other engineered safeguards and nuclear safety l systems, is to limit the release to the environs of radioactive materials so that offsite doses from a postulated DBA will be below the guideline values of 10CFR100. 5.3.2' Safety Design Basis The safety design bases of the SCS are as follows: 1. The SCS shall be designed to provide secondary containment when the primary containment is operable and when the primary containment is open i 2. The SCS is designed with sufficient redundancy so that no single active system component failure can prevent the system from achieving its safety objective 3. The SCS shall be designed in accordance with Class I design criteria. (Exception to this is the containment access locks. Since simultaneous LOCA's and SSE's are not ~~ postulated, the access locks shall be designed in accordance with Class II design criteria. The access lock door lying directly in the SCS shall be designed in accordance with { I Class I design criteria so that the possibility of a ground i level release to the environs through the access locks is j eliminated if a seismic event were to follow or precede an accident which results in a contaminated reactor building atmosphere.) The SCS is not designed to withstand torrado loads ~ 4. The secondary containment shall be designed to limit the ground level release to the environs of airborne radioactive materials so that offsite doses from a design basis fuel handling, or loss of coolant accident (LOCA) will be below the guideline values stated in 10CFR100 5. The Reactor Building shall be designed to contain a positive internal pressure of at least 7 in of water 6. The SCS shall be designed to be sufficiently leaktight to allow the Standby Gas Treatment System (SGTS) to reduce the Reactor Building pressure to a minimum subatmospheric pressure of 0.25 in of water, under neutral wind conditions, when the SGTS fans are exhausting Reactor Bui.iding atmosphere at a maximum of 4,000 ft'/ min / 7. The Reactor Building Isolation and Control System (RBICS) shall be designed to isolate the Reactor Building sufficiently fast to prevent fission products from the ( postulated fuel handling accident from being released to the 5.3-1 Revision 6 - July 1986

PNPS-FSAR environs through the normal discharge path 8. The SCS is provided with means to conduct periodic tests to verify system performance i. ^ u b 5.3-la Revision 6 - July 1986

PNPS-FSAR 5.3.3 Description 5.3.3.1 General The SCS consists of four subsystems. These subsystems are the Reactor Building, the RBICS, the SGTS, and the main stack. The SCS surrounds the Primary Containment System, and is designed to provide secondary containment for the postulated LOCA. The SCS also surrounds the refueling facilities and is designed to provide primary containment for the postulated refueling accident. The SCS utilizes four different features to mitigate the consequences of a postulated LOCA (pipe break inside the drywell) and the refueling accident (fuel handling accident). The first feature is a negative pressure barrier which minimizes the ground level release of fission products by exfiltration. The second feature is a low leakage containment volume which provides a holdup time for fission product decay prior to release. The third feature is the removal of t particulate and iodines by filtration prior to release. The fourth ~ feature is the exhausting of the secondary containment atmosphere through an elevated release point which aids in dispersion of the effluent by atmospheric diffusion. Each of the features is provided by a different combination of subsystems: the first by the Reactor Building, the RBICS, and the SGT exhaust fans; the second by the Reactor Building and the RBICS; the third by the SGTS filters; and the fourth by the main stack. 5.3.3.2 Reactor Building The Reactor Building completely encloses the reactor and its pressure suppression Primary Containment System. The Reactor Building houses the refueling and reactor servicing equipment, new and spent fuel storage facilities, and other reactor auxiliary and service equipment. Also housed within the Reactor Building are the CSCS, Reactor Cleanup Demineralized System, Standby Liquid Control System (SLCS), Control Rod Drive (CRD) System, Reactor Protection System (RPS), and electrical equipment components. The structural design features of the Reactor Building are described in Section 12. Discussions of the Reactor Building's Class I design are included in Section 12 and Appendix C. The Reactor Building '- also designed to meet the shielding requirements discussed in Section 12. 5.3.3.3 Reactor Building Isolation and Control System The RBICS Terves to trip the Reactor Building supply and exhaust fans, isolate the normal ventilation system, and provide the starting signals for the SGTS in the event of the postulated LOCA inside the drywell, or the postulated fuel handling accident in the Reactor Building. Either of two signals will initiate the SCS. These ' signals, which indicate a LOCA inside the drywell, are high drywell pressure or low reactor water level. In addition, radiation monitors [ in the operating (refueling) floor ventilation exhaust duct, which 1 5.3-2 l ~~ ~~

PNPS-FSAR Indicate a fuel handling accident, can initiate the SCS. Secondary containment can also be initiated manually from the control room. Normally open, air-operated isolation dampers are provided on the discharge side of the Reactor Building and operating floor supply fans. Normally open, air-operated isolated dampers are provided on the intakes to the operating floor ventilation exhaust fans, the clean area exhaust fans, the contaminated area exhaust fans (upstream j of the filter assemblies), and the control rod drive maintenance room exhaust fan. See Figure 5.3-1. Two dampers in series are provided throughout the isolation system to provide the required redundancy. Both dampers fail closed upon loss of de power to the solenoids or upon loss of instrument air to the dampers. The isolation dampers are piston operated and designed to close within 3 sec after receipt j of the secondary containment initiation signal. Penetrations of the secondary containment are designed to have leakage characteristics consistent with secondary containment leakage y requirements. Electrical penetrations in the Reactor Building are designed to withstand normal environmental conditions and to retain their integrity during the postulated fuel handling accident and the LOCA inside the drywell. Two interlocked sealed doors on the equipment and personnel access locks assure that building access can not interfere with maintaining the secondary containment integrity. - - All normally open drains which are open both to the secondary containment and the outside atmosphere are provided with water seals to maintain containment integrity. This is exemplified by the four 14 in dewatering lines for the reactor auxiliary bay floor sumps. These lines penetrate the secondary containment boundary, two below each of the two sumps, and terminate in a pair of troughs within the i torus compartment. The two 4 ft cubic shaped troughs, located adjacent to the east wall, maintain containment integrity by providing water seals for each of the f.our lines. High and low levels are alarmed in the control room. On low level, the operators are directed by procedure to refill the

troughs, to ensure containment' integrity.

5.3.3.4 Standby Gas Treatment System The SGTS consists of

two, identical, parallel air filtration assemblies separated by an 18 in thick concrete block wall and completely enclosed within a

Class I structure. Each of the filtration assemblies are full capacity. Each consists of a demister, an electrical heating coll, a high efficiency particulate absorber (HEPA), two charcoal filter beds, and a final HEPA filter. With the Reactor Building isolated, each of the two fans has the necessary capacity to reduce and hold the building at a minimum subetmospheric pressure of 0.25 in of water. Each fan has a design flow rate of 4,000 f t'/ min. Motor-operated exhaust fan outlet damper controls are provided to maintalh the required negative pressure. See Figure 5.2-17. The system includes ) isolation dampers which fail open on loss of dc power to the l l l 5.3-3 1

PNPS-FSAR I* solenoids or upon loss of instrument air to the air operators on the ( dampers. The demister is designed to remove entrained water droplets and mist from the entering air stream. The electric heating coil is designed to reduce the relative humidity of the air stream to 70 percent. An i interlock with its associated exhaust fan prevents the heating coil I from operating when the fan is snut down. Each HEPA filter is designed to be capable of removing at least 99.97 percent of the 0.30 micron particles which impinge on the filter. The charcoal filters are iodide-impregnated activated carbon filters capable of removing in excess of 99 percent of the iodine in the air stream with 10 percent of the iodine in the form of methyl iodide (CH I) under 3 entering conditions of 70 percent relative humidity. l The accident evaluations using the. standard NRC appr$ach are described in Section 14.9. In these analyses the SGTS charcoal filters were credited with removal of 95 percent of the influent i iodine. The system will start automatically upon a high radiation signal from the operation (refueling) floor ventilation exhaust duct monitor, or upon receipt of high drywell pressure or low reactor water level signals. The system can also be manually started from the control room. Upon receipt of any of the initiation signals, both fans start, all SGTS isolation dampers open and each fan draws air from the isolated Reactor Building at a flow rate of approximately 3 ~ 4,000 ft / min. After a preset time delay, one fan is stopped. Cross-connections between the filter trains are provided to maintain the required decay heat removal cooling air flow on the charcoal filters in the inactive treatment train. The system discharges to the main stack through a 20 in underground line. The SGT fans are powered from the emergency seTvice portions of the auxiliary power distribution system. Drywell and torus purge exhaust can also be directed to the SGTS for processing before release up the main stack. See Section 5.2. The High Pressure Coolant Injection System (HPCIS) gland seal steam condenser exhauster discharge is also routed to the SGTS during accident conditions. The Reactor Building Heating and Ventilating System is discussed in Section 10.9. 5.3.3.5 Main Stack i The location of the main stack is shown on Figure 1.6-1. The top of the stack is at elevation 400 ft msl. The structural design of the stack is discussed in Section 12. 5.3.4 Safety Evaluation The SCS provides the principal mechanisms for the mitigation of the ~ consequences of an accident in the Reactor Building. The primary and secondary containment act together to provide the principal mechanisms for the mitigation of the consequences of an accident in 5.3-4 Revision 7 - July 1987

PNPS-FSAR the drywell. If the leakage rate of the building is low, and the leakage air is filtered and discharged to the elevated release point I (utilizing the SGTS and the main stack) the offsite radiation doses that result from postulated accidents are reduced significantly. The Reactor Building is a Class I structure (with the exception of the secondary containment access locks which are Class II structures) designed in accordance with all applicable codes. Desi n of the Reactor Building for a maximum inleakage rate of 4,000 ft / min at a building subatmospheric pressure of 0.25 in of water at neutral wind conditions, results in a low exfiltration rate even during high wind conditions. In the event of a pipe break inside the primary containment or a fuel handling accident, Reactor Building isolation will be effected and the SGTS will be initiated. Both SGTS exhaust fans will start. After a preset time delay, one fan is stopped. With the Reactor Building isolated, each fan in the SGTS has the capability to hold the building at a subatmospheric pressure of 0.25 in of water when drawing air from the building at a flow rate of 3 4,000 f t / min. Exhaust fan outlet damper controls on each fan are -r* provided to maintain the required flow rate. The RBICS performs the required isolation actions of the SCS following receipt of the appropriate initiation signals. Following I initiation, the Reactor Building ventilation isolation dampers close within 3 sec. The RBICS also automatically trips the Reactor Building supply and exhaust fans, and starts the SGTS. The normal design flow rate in the Reactor Building operating (refueling) floor 3 exhaust duct is 40,000 ft / min. During shutdowns, the flow rate is increased to approximately 50,000 f tJ/ min at which time it takes more than 3 set for fission products released in any postulated fuel handling accident to travel from the operating (refueling) floor ventilation exhaust radiation' monitors to the i solation dampers. Thus, no direct release of fission products to the environment (bypassing the SGTS filtration processes, and the elevated release point provided by the main stack) is possible. The SGTS filters exhaust air from the Reactor Building and discharges the processed air to the main stack. The system filters particulate and iodines from the air stream in order to reduce the level of airborne contamination released to the environs via the main stack. When the system is exhausting from the Reactor Building, the building is held at a minimum subatmospheric pressure of 0.25 in of water. l l Appendix G identifies requirements for establishing secondary j containment (Safety Action 27), following an assumed pipe break i inside the primary containment (Event 39), and following an assumed spent fuel handling accident (Event 40). Secondary containment is not established following assumed pipe failures which result in the release of steam into the Reactor Building (Event 41). l The following piping which is located within the Reactor Buildin'g and normally contains hot fluids at reactor pressure was considered: High Pressure Coolant Injection (HPCI) turbine steam supply line; 5.3-5 Revision 6 - July 1986

PNPS-FSAR Reactor Core Isolation Cooling (RCICI turbine steam supply line; and Reactor Water Cleanup System (RWCU) piping. J J The maximum rate of steam release into the Reactor Building and the I corresponding period of steam release was calculated for the above piping: 1. HPCI steam line: Maximum release rate, 300 lb/sec of steam Period of release, 22 sec 2. RCIC steam line: Maximum release rate, 25 lb/sec of steam Period of release, 17 sec j 3 RNCU piping-i Maximum release rate, 250 lb/sec of steam i Period of release, 22 sec Steam leakage into the Reactor Building could be exhausted through the ventilation exhaust systems operating at normal building pressures at a calculated rate of 63 lb/ set of steam. The SGTS 1 operating at normal Reactor Building pressures could exhaust about l 5 lb/sec of steam. Steam leakage in excess of these amounts would . _ _ result'in Reactor Building pressure increases above normal. The Reactor Building is designed to relieve excessive internal / pressures so as to preserve main structural integrity, considering the rapid pressure reduction outside the building associated with tornadoes. Refer to Appendix H.5. Reactor Building differential pressures exceeding about 0.5 psi will be relieved through the ) Reactor Building roof. Steam leakage within the normal operating capability of the { ventilation exhaust systems would be ducted to the main building exhaust vent. The ventilation exhaust from the principal Reactor Building compartments housing the RCIC steam supply line and turbine, the HPCI steam supply line and turbine, and the RWCU are monitored by temperature elements. These elements provide temperature indication and high temperature alarms in the main control room. The temperature set points for these alarms will alert the operator to potential steam leakage conditions at leakage rates that are less than the normal operating capability of the ventilation exhaust systems. Steam leakage rates that exceed the capability of the normal Joperating ventilation exhaust systems could result in abnormal ventilation flow paths and abnormal Reactor Building extaust locations. The design of the Reactor Building would indicate that likely abnormal release locations would include the building roof and - building access locations. Steam leakage into a compartment within the operating capability of ( the ventilation exhaust systems would be confined within the normal 4 5.3-6

PNPS-FSAR exhaust

paths, and therefore would limit the steam-flooding principally to the compartment where the leakage originated.

Thus the operability of safety related equipment,

controls, and instrumentation located in other compartments would be maintained.

The ventilation exhaust temperature sensors will detect steam leakage from the RCIC steam line, the HPCI steam line, or the RWCU piping at leakage rates that are below the normal operating capability of the l ventilation exhaust from the compartments housing these

hot, pressurized lines.

Early detection of steam leakage at rates below the capability of the normal ventilation systems and subsequent isolation of leaks provide protection of safety related equipment within the Reactor Building. See Section 7.3. The main stack provides an elevated release point for airborne activi* ' during the postulated station loss of coolant and refueling accidents. Release of activity to the environs from the Secondary Containment System is analyzed in detail in Section 14 Station Safety Analysis. It is concluded that the safety design bases are met. 5.3.5 Inspection and Testing The secondary containment leakage rate can be determined.i n the ~ 1 following manner. The Reactor Building is isolated and the SGTS is started with one treatment train and its associated exhaust fan. The - exhaust flow rate is controlled by the fan outlet damper control position as determined by flow rate measurements in the SGTS exhaust duct. The fan outlet damper positioner is used to control the ) exhaust flow rate at 4,000 ft'/ min. If the subatmospheric pressure as measured within the Reactor J Building is equal to or exceeds 0.25 in of water (with neutral wind conditions at the site) the building safety design basis leaktightness with respect to inleakage is verified. Tests of the ability of the various isolation initiation signals to automatically render the Reactor Building isolated, to trip the supply and exhaust fans, and to start the SGTS can be conducted by simulating the isolation signals. Provisions are made for periodic tests of each filter unit. These tests include determinations of differential pressure across each filter and of filter efficiency. Connections for testing, such as injection and sampling, are located to provide adequate mixing of the injected fluid and representative sampling and monitoring, so that test results are indicative of performance. Each HEPA is tested with 00P (Di-octyl-phthalate) smoke. The charcoal filters can be _ tested ~ for bypass with freon. J The electric heating coil in each filter train is tested to show that the relative humidity of an entering air stream is reduced. l i 5.3-7

+

• e PNPS-FSAR bkt 5.3.6 Proposed Nuclear Safety Requirements for Initial Plant Operation General The entries in this section represent the proposed nuclear safety requirements for the SCS for each BWR operating state which r.epresents an extension of the stationwide BWR systems analysis of Appendix G.

The following referenced portions of the Safety Analysis Report provide important information justifying the entries in this section: Reference Information Provided 1. Earlier parts of Section 5.3 Description of the SCS 2. Station Nuclear Safety Opera-Identifies conditions and tional Analysis Appendix G events for which the SCS ~ is required 3. Jacobs, I.H. Guidelines for Describes methods used to Determining Safe Test Inter-establish allowable repairs vals and Repair Times for times Engineered Safeguards. General = (lectric Co., Atomic Power Equipment Department, APED - 5736, April 1969 Each detailed requirement in this section is referenced, if possible, ) to the most significant station condition originating the need for the requirement by identifying a matrix block on one of the Matrix 3 sheets of Table G.5-3. The matrix block references are given in parentheses beneath the detailed requirements in the " minimum ~ required for action" section. The matrix block references identify the BWR operating state, the event number, and the system number. For example, F40-91 identifies BWR operating state F (Matrix 3), event (row) No. 40, and system (column) No. 91. System Action The SCS operates to limit the release of airborne radioactive materials to the environs. Number Provided by Design 1. One Reactor Building 2. One main stack 3. One RBICS, with two dampers in series provided throwghout the isolation system to provide redundancy. The control f system is designed such that all dampers fail closed on loss a 5.3-8

PNPS-FSAR U$ i of dc power to the solenoids or on loss of instrument air to tne valves. 4 One SGTS consisting of two identical, parallel air filtration assemblies and two full capacity exhaust fans. Minimum Required for Action SWR Operating States A,B.C.D.E, and F: e, The Reactor Building (A40-90) (B40-90) (C39-90) (D39-90) (E39-90) (F39-90) The Main Stack (A40-105) (840-105) 3 (C39-105) (D39-105) t (E39-105) (F39-105) One damper at each isolation point (with associated-controis) (A40-102) (840-102) (C40-102) (40-102) (E40-102) (F40-102) One flitration assembly train and one exhaust fan (A40-91) (B40-91) (C39-91) (039-91) (E39-91) (F39-91) Proposed Condition Required for Continuous Oceration Limiting Condition for Operation The Limiting Conditions for Operation are set forth in the Technical Specifications referenced in Appendix B. Proposed Surveillance Requirement for Initial Plant Operation The active components of the SCS shall be tested once every 3 months. The filter efficiencies will be tested once every 12 monthi. 5.3.7 Current Operational Nuclear Safety Requirements The current limiting conditions for operation, su rvel'11ance requirements, and their bases are contained in the Techni cal Specifications referenced in Appendix B. 5.3-9 _ _ _ _ _ _ _ _ -}}