ML20024C664
| ML20024C664 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/11/1983 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| References | |
| TASK-*, TASK-GB GPU-2049, NUDOCS 8307120937 | |
| Download: ML20024C664 (25) | |
Text
.__
? f. "
3 N g& Ed.2o99 6.3 CERGE' ICY CORE COOLING SYSTEM
!O ne high-pressure injection, icv-pressure injection, and core flooding sys-te=s collectively form an e=ergency core cooling system (ECCS). This system s
furnishes cooling water to the reactor core to compensate for a loss of nor-mal cooling capacity that could result from a loss-of-coolant accident (LOCA).
6.3.1 DESIGN BASES te principal design basis for providing protection over the entire spectrum of break sizes for the ECCS is AEC General Design Criterion 35 Very small breaks that do not actuate the engineered safety features mode of operation vill be accce=edated by the normal makeup systes ss. required by AEC General Design criterion 33. Separate anq independent fiev paths are provided in the v
ECCS, and redundancy in the active cc=penents ensures that the required func-tions will be perfor=ed if a tingle failure occurs. Separate e=ergency pcVer sources are supplied to the redundant active components, and separate instru-ment channels are used to actuate the systems. Actuation pressures for the ECCS are given in Tech. Spec. 3 5 3 6.3.1.1 Rans:e of Coolant Ruutures and Lesks I
. De emergency core cooling systes (ICCS) is designed to mitigate the conse-quences of all breaks of the reactor coolant system pressure bounda:y which result in loss of reactor coolant at a rate in excess of the capability of P,
the reactor ecolant =akeup system up to and including a break equivalent in area to,the double-ended rupture of the largest pipe of the reactor coolant
! U system.
sis has been performed for a spectrum A multinede icss-of-coolant (LOCA) angft ) and was based on analytical tech-2 to ik.1 of break sizes (fres 0.5 ft niques, assu=ptions, and procedures similar to these found in part k, A; pen-dix A, of the AEC Interim Acceptance Criteria for E=ergency Core Cooling Sys-tems for Light 'Jater Pcuer Reactors, dated June 29,(1ST1 as a= ended on Dece -
The multinode bicvdown code - CRAFf, kJ RITLCOD,(7 and the cladding heatup code - TEAL-3,(6)the reflooding code -
ter 18,1971.
are used to perform these analyses at a core pcuer level of 2772 MWt. Specific discussion ci the consequences of each break along with curves showing the pars =eters of intere'st are presented in 6.3.3.2.
Se evaluation of loss-of-coolant accidents re-sulting frca s=all breaks (less than 0.5 fe ) in the reactor coolant system z
was also performed and is reported in 6.3.3.3.
The results of the above analyses, which cover the full spectrum of postulated icss of coolant accidents, desenstrate that the e=ergency core ecoling systes vill terminate the cladding temperature transient and limit the course of the LOCA in accordance with the final Acceptanca Criteria.
l 6.3.1.2 Fissica Product Decay Heat l
The decay heat curve described in the prepesed ANS Standard (approved by sub-ccanittee ANS-5 of American Nuclear Society on June 11,1968) with a +205 8307120937 830711 Pun Aoucn 050002e9 6.3-1 Am. 51 (2-4-77) l P
HOL
.e occur when high RC system pressure is maintained and delays uncovering.of the core when intermediate-sized Isaka occur. LPI and the CT system inject bo-rated water into the core at intermediate-to-low RC system pressures and en-sure adequate core' cooling for break sizes ranging from intermediate to a break equivalent to the double-ended rupture of the RC piping in either the hot or the cold leg.
Af ter draining the BWST to the lov level point, suction is switched to the reactor building sump. The LPI system recirculates the spilled reactor coolant and injection water from the reactor building sump to the react,r o
vessel and maintains long-term core cooling.
For small breaks the RC system pressure may be higher than 'the maximum decay heat removal pump head at the time of iesctor building emergency water re-Under these circumstances a crossover connection vill permit circulation.
alignment of the makeup pumps to take suction from the outlet of the decay heat removal coolers to provide for sump water recirculation to the reactor core.
In sunmary, ECCS and related pumps that must operate following a LOCA include the HPI pumps and the LPI pumps (which take suction from the BWST) and the CF system during the initial injection phase. When the BWST is exhausted, the LPI pumps take suction from the reactor building sump and provide recirculation throush the DE removal coolers to the reactor vessel, thus pro-j viding long-term ccoling of the core.
.O 1
-O 6.3-3 Am. 51 (2-4-77)
~
, ~.,
The 1PI lines and components are designed for normal reactor cooldown operating conditions since they are part of the. DH removal system. These system pressure and temperature requirements are greater than those encountered during ECCS operation. The HPI lines and components vill be designed for the normal oper-ating conditions as part of the makeup system. Although the HPI lines do not serve any function during normal reactor operation, they were designed as part of the makeup system since the makeup system piping and valves are sub-jected to more severe conditions during normal operation than during emergency operation.
6.3.2.6 Coolant Storage The capacity of each of the coolant storage facilites required by the ECCS is shown in Table 6.3-1.
6.3.2 7 Pumo Characteristics The total dynamic head, NPSE and brake horsepover are shown in Figure 6.3-2 for the makeup pumps used for KPI and in Fir.ure 6.3-3 for the DH renoval pumps used for LPI.
6.3 2.8 Neat Exchanger Characteristies The IJI coolers are designed to remove decay heat generated during a normal shutdown. In addition, each cooler is capable of cooling the injection water during the recirculation mode following a loss of coolant accident. The heat transfer characteristics of the low pressure injection cooler when cooling the O
core by recirculating the water from the resctor building sump are given in Table 6.3-1.
6.3.2 9 ECCS Flow Diagrams The flow diagram of the ECCS is shown in Figure 6.3-1 with node points located to describe emergency and test conditions. The temperature, pressure, and flow rates at these nodes are given in Table 6.3-4.
l 6.3.2.10 Relief Valves and Vents The capacities, settings, and functions of the relief valves in the emergency core cooling system are listed in Table 6.3-5
~
6.3 2.11 System Reliability _
System reliability is assured by ecepliance with the intent of the AE,C General Design Criteria and by the system functional design including equipment re-dundancy, separation of redundant equipment, use of normally operating equip-ment for safety functions, equipment testability provisions. System reliability is also assured by proper component selection, by physical protection and arrangement of the system, by using proven ccaponent designs wherever possible and/or conducting tests, and by quality control and assurance requirements i
implemented during the design, manufacture, and installation of the components and systems.
b The HPI and LPI systems each consist of two separate and redundant injection strings. Each string of the two HPI strings can supply 100% of the design 63 3
-5 _,__
.a 1ayout of the systess, sufficient clearances between.the cesponents of the system and connecting systa=s and an appropriate-systes analysis.
I
- A Protection against =issile da= age is provided by direct shielding or physical separation of redundant equip =ent.
The =ajor active ccaponents of the ECCS are external to the R3 and therefore are not ex;osed to the post-accident R3 en-viron=ent. Since most of the ECCS piping within the R3 is located outside the primary and secondary shield, it is protected from missiles, originating frca these areas. ECCS piping that is not fully protected against LOCA missile damage, utilizes dual lines to preclude loss of the protective function. Missiles that may be generated in one loop cannot rupture ICCS lines in the other loop.
The EP Lines enter the reactor building through penetrations in different sec-tions of the building. Each injection line splits into two lines outside the reactor building to provide four injection paths to the RC system. The four connections to the RC system are located between the reactor coolant pump dis-charge and the reactor vessel inlet nozzles. Four injectics lines penetrate the missile shields so that the effect on injection flow is minisized in the unlikely event of missile damage to an injection line inside the secondary shield.
Protection from missiles is previded for the low pressure injection lines within i
the reactor building. Me portion of the low pressure injection system located in the reactor building consists of two redundant injection lines which are 1
connected to injection nozzles located on opposite sides of the vessel. Both redundant suction lines from the sump are missile protected. The sump suction is located o.utside of the secondary shielding and is additionally protected by a grating.
Se entire chre flooding system is located within the reactor building. The
[
core flooding tanks and two of the three valves in each core flooding line are
~
l-located outside of the secondary shield.
6.3 2.13 Provisions for Perfor=ance Testing Provisions have been ine'orporated in the ECCS to allow performance testing of the active system ccaponents concurrent with unit operation. The low pressure injection system design incorporates a test flow path to demonstrate the opera-bility of the Decay Eest Removel pumps. The flow paths take flow from, and re-turn it to the BWST. The testing procedure involves opening the BWST outlet valves,'and the test line valves.
Se high pressure injection pumps can be tested by running each pump and taking suction from the makeup tank. The discharge vill flow through the recirculation lines back to the makeup tank.
The core flooding systaa has no active components that require periodic testing to assure reliability.
6.3.2.lb Net Positive Suetieri Head The NPSI required by the ECCS Pumps are as follows:
Minimus Available NPSH. Feet Recuired NPSE. Ft.
Decay Heat /LPI Pumps See Table 6.2-11 O
Makeup / EPI Pumps (Injection
%w' phase, suction from BWST) 29 0 26.0 Makeup /HPI Pumps (Recirculation phase, suction from Decay Heat /
LPI Pmups) h3.279 29 39
.L
~.
6.3.2.16 Motor-coerated valves and Controls Coritrol of the decay heat removal system is required for nor=al cooldown of the I
unit and for low pressure injection for emergency core cooling. There are two redundant flow loops provided. Each loop contains a pump, a reactor coolant inlet valve, a BWST succion valve, and a reactor building sump suction valve.
An of these components can be controlled by the operator from the control room.
Each of the high pressure, actor-operated valves in the suction line from the reactor coolant system to the decay heat pumps has independent controls. These controls are designed to close the valves automatically or to prevent the opening of the valves when the reactor coolant pressure is above the design pressure of the decay heat suction piping. This prevents overpressurizing the decay heat system in the event the valves are inadvertently left open during heacup or an operator tries to open the valves prematurely during cooldown.
In the event of an accident requiring safety action, the safety features actuation system will actuate both pumps, both reactor coolant inlet valves, and both BWST suction valves. When the water level in the BWST decreases to the predetermined low-low level point,.the reactor building sump succion valves automatically open to provida continuous, long-tera emergency core cooling.
The controls in this discussion comply with the intent of IEEE-279-1971. Each inattument loop for the redundant pumps and valves is independent. This redundancy and independence assures that any single failure will not prevent the system from performing safe shutdown. The redundant channels are electrically independent and physical separation is provided. All instrumancation is designed to withstand the f"'N range of environmental conditions expected and any energy supply variation foreseen.
V Manual operation of all systems can be accomplished from the control room or locally at the equipment. The system is designed to facilitate repair, replacements, and adjustment of components and modules. Wiring for the separate channels is color-coded for identification and differentiation.
6.3.2.17 Manual Actions t
At the completion of the automatic transition from initial injection to recirculation, I the operator is required to open the HPI-LPI crossover valves from the control room to provide a suction supply to the HPI pumps long term. The indications the operator must watch are located on panels 8 and 15 in the control roca, and are clearly visi-ble from the normal control station. The crossover valve handswitches are also located in the control room on panel 8 (see Figure 7.5-1 for control room panel arrangement). There are no permissive interlocks between ECCS components that could jeopardise the overall system performance. The limiting accident requiring this action is the core flood line break in combination with the single failure of the opposite train diesel generator, and loss of off-site power. The operator has greater than 120 minutes to complete chils action from the start of the accident, and greater than 60 minutes from the start of the recirculation phase, since a significant amount of available water for the HPI pumps remains in the BUST after the switcho'ver to the sump. The operator will be required to complete this action for all loss of coolant accidents, thereby elh ing the need for break location identification.
t v'
6.3-sa Am. 63 (2-10-78)
(
6.3 2.15
? recess Instrumentatien l
Tr., fellowing process instru=entation is available in the control room to cssist the cperator in assessing post LOCA conditions: RC loop and unit out-le: and inlet temperature, RC loop and unit average temperatures, RC loop and ur.it, te=perature differences, RC Ic5p pressure, pressurizer level, pressurizer te=,.erature, RC loep flow, RC total flow, containment pressure, core flooding i
,ta:2 level and pressure, HP injection flow, decay heat removal cooler inlet te=;erature, decay heat removal discharge pressure, decay heat removal flor, and borated vater storage tank te=perature and level.
In addition to the above process instrunentation, the following equipment or systa= status is annunciated in the control room:
n.
Centain=ent Iselation Valve Status Valve position (open or closed) is indicated by lights and a computer print-out sh:nri g valve position is available.
b.
STAS Status Indicating lights previde the following infor=atier. for each emergency core.ccolant injection channel:
1.
Channel bypass permit.
(
2.
Cha= e1 typassed.
3.
3ypass reset permit.
k.
3ypass reset.
5.
Safety features histable tripped.
6.
Protective function fully enabled.
Peactor 3uildin: Iselation and Coolice Status e.
I dicating lights provide the following information for each channel:
1.
Safety feature histable tripped.
2.
Prctective function fully enabled.
3 Channel defeated.
h.
Cha==e1 defeat reset.
5.
Post-actuation defeat permit, d.
Safetv Teatures Eeui :=ent Status Indicat*.ng lights provide the following information for equipnent required opera.e during and after an accident:
6.3-10
... u
., =.
O 1.
Pu=ps-running, not running.
2.
Valves - open, closed.
In addition, ECCS low flow probes and indicators provide local indication to ensure sufficient flow to prevent a high boric acid concentration in the core region.
~
6.3.2.19 Maeerials The materials used for the ICCS components are given in Table 6.3-1.
All active components are located outside the reactor building and therefore are not required to operate in the post-T.0CA reactor building environment.
See 6.2.2.2.1.3 for a discussion of materials considerations for components inside the reactor building.
1 l
m.
6.3-11 Am. 48 (11/13/76)
' j
~ ~ -
-v-n.--,n._
_,w._,. _.,, _,,., _ _., _, _ _ _,,,. _ _ _, _ _, _ _ _ _ _
b) If an ISTA signal occurs while two pumps (1A & 13 or 13 & 1C) are running, t
q the la pump will be tripped immediately. A time delayed trip signal is supplied to the other running pump in the event that the 13 pump fails to trip.
4 c) For either case a) or b) the non-operating pump on the makeup header will remain as a standby pump and cannot be started while the other pump is operating. This lockout is accomplished by a breaker con-tact from the non-running pump which is closed in the running pump control circuit thus preventing a falso trip of the running pump.
d) Should either E-P-1A or IC be taken out of service, W-P-13 would be manually valved to the sans section of the common suction and discharge headers and manually transferred to the same power supply as the out-of-service pump. Both pumps would be started by an ISTA signal.
In no instance would a single. failure prevent either minimum ECCS capability, or allow more than two pumps to run (one on the makeup header, and one on the BWST header). During normal operation, provision is made to allow the running of two makeup pumps (IA & 13 or 13 & 1C) on the sa=e bus while drawing suction from the makeup tank. This operating moda is required for pump switching to maintain con-tinuous flow to the RC pump seals, and for periodic testing of the ECCS and makeup pumps themselves. Af ter DR-V5A/B have opened completely, W-V12 will be closed manually, terminating suction from the makeup tank.
As reactor coolant pressure decreases, the discharge flow from the makeup pumps will increase. When the reactor pressure has decreased to approximately 200 psig.
I]
the decay heat removal pumps are capable of providing the required injection.
V Supply to the EPI system for the initial 10 seconds post 10CA will be solely from the makeup 'cank until valves DE-VSA/B open. With makeup pump operating at runout flow of $50 spa, 91.7 gallons of water would be drawn from the makeup tank during 1
this 10 second period. Assuming the makeup tank is at the worst allowable level l
and pressure allowed by operating procedures when the accident occurs, 15 inches of i
water remain in the tank after draudown.
I In order to consider the worst possible consequences, the 10CA is assumed to occur j
in the core flood line in the saaa safety train as the operating makeup pump, and no loss of power occurs. The operating makeup pump continues to run and the Es.
makeup pump (suction from BWST) starts. Both pumps are assumed to run out to 550 gym instantaneously. The decay heat pump on the side of the break starts and runs out to 4150 spa (4060 spa to the break and 90 gym recirculation). The other decay heat pump starts and feeds the core through the intact core flood line at 3200 spa rated flow. Both Reactor Building spray pumps start and operate at assumed conser-vative runout conditions of 1760 and 1740 spa.
Based on the above considerations, the line from the makeup tank would have a mini-uma of two feet of water above the main suction header at all times, hence, hydrogen would not be drawn into the system. As added precaution, operating procedures re-guire the operator to close M -V12, thereby isolating the makeup tank as quickly as possible after DE-VSA/B have opened.
e e
6.3-12a Am. 65 (.5-11-781
vessel; (3) the KPI system nay be operated through the ipI during recirculation if injection through the EPI pu=,;s is required, at that tine.
]
Core Floodinx System The injection response of the core flooding systen depends on the rate of re-duct', of the reactor coolant system's pressure. The core floodi=g nozzlas and lines are designed to ensure that they can accom=odate the differential temperature occurring between the injection mode and the recirculation mode.
6.3.3.1 Evaluation Model The e ectiveness of the energency core cooling system (ICCS) in the event of an RCS piping failure, has been evaluated and shown to comply with the five acceptance criteria of 10 CFR 50.h6.
Cceformance is de=enstrated by analyses conducted under the general guidelines of Appendix E to 10 CFR 50, as interpreted by the 2&W evaluation model documented in RAW 10104.
6.?.3.2 ICCS Perfor=ence The evaluation of the 3shcock & Wilcox 1TT-Fuel Assembly lovered loop nuclear steas systen during a hypothetical loss-of-cociant accident (LCCA) is presented j
in BAW10103. This report describes the application of B&W's evaluation model, with appropriate sensitivity studies to select certain features of the model,
- to evaluate the consequences of a spectrum of hypothetical ICCAs, and to
- determine allevable linear heat rates as a function of elevation in the core for the most limiting break. Calculations are also presented to descestrate that
~
the TMI-2 NSS complies with the five acceptance criteria of 10 CFR 50.h6 for 1
ICCSs. In brief, the five acceptasi:e criteria and a general overview of the compliance presented in RAW 10103 are as follows:
1.
The peak cladding temperature shall not exceed 2200F. Compliance: A spectrum of breaks is evaluated, and allevable linear heat rates as a function of core elevatics for the most limiting tresh are determined.
These results, presented in Sectiens 6 and 7 of BAW10102, fors ; art of the data tase frcs which administrative controls and procedures durir.g pcver operation are established. The conservative nature of the evaluation nodel ecsbined with the imposed restrictions during power operation ensure the effectiveness of the ICCS to id-4t cladding tem;eratures to values less than 2200F in the unlikely event of a LOCA.
2.
The percentage of local cladding oxidation shall not exceed 175 Compliance: The analysis perfor-sed to satisfy the first criterion (Sections 6 and T cf 3AW10103) also provides the percentage of local cladding that oxidizes. These results show that oxidation of,the hot pins is less than 175.
3.
The percentage of hydrogen generation resulting frca whole-core c1=Ad4 = cxidation shall not exceed 15.
Compliance: A separate 9
I 6.3-13 Am. 39 (3/30/76)
I
.._-.-._.___-.-__m.,.._..---,,m._____,-.__--,-.,,mm,
.y__
_,__.-__.--..m-,,,-
I
=
e.
O q
the maximum linear heat rate covered by BAW-10103, the transient cladding surface temperature will never exceed its initial value of 660F, no metal a
water reaction will occur, and the core geometry will remain cociable as no cladding rupture will occur. Iang term cooling is established as the -
HPI and LPI pumped injection systems provide fluid in excess of the boil-off rata due to core decay heat.- Thus the five Acceptance Criteria ir.
10CFR50.46 are met.
The new modeling techniques used in the CRAFT 2 analyses for the present studies show improvemen,t in the core performance when compared with the results of the same breaks reported in BAW-10052 and BAW-10064. Therefore, if all the breaks reported in BAW-10052 and BAW-10064 were re-analyzed with the present model, the same trend of improvement in core performance would be realized. Thus, the present analyses in conjunction with the analyses of BAW-10052 and BAW-10064 provide a suitable nail break spectrum for demonstration of compliance of the ECC system with the five Acceptasca Criteria in 10CFR50.46.
1 The analysis uses the CRATT2 code to develop the history of the reactor coolant system hydrodynamics. For sus'1 leak analysis it is sufficient to use smaller models than are used for irge loss-of-coolant studies because hydrodynamic responses are slow enough for simpler models to describe them.
The CRAFT model uses 19 nodes to simulate the reactor coolant system, two nodes for the secondary system, and one node for the reactor building.
l Control volumes (nodes) in and are;ad the vessel are all connected by a pair of flow paths to allow the occurrence of counter-current flow. The eQ break is located in the cold leg piping either at the lowest point in the pipe at the pump suction, or at a point opposed to the high-pressure injection I
nozzle at the pump discharge, or in the core flood line joining the CF nozzle.
The Wilson, Grenda and Patterson average bubble rism model is used for all nodes. Within the core ragica, however, a multiplier of 2.38 is applied to the calculated bubble rise velocity. This report demonstrates that a multi-plier of 2.38 in CRAIT2 gives a mixture height within + 2% of that predicted by 70AM. Thus, no FOAM analysis will be needed if the CRAFT 2 mixture level remains above the core by 2% of the active length.
The following assumptions are made for conditions and system responses during the accident:
1.
The reactor is operating at 102% of the steady-state power level of 2772 Mwt.
2.
The leak occurs instantaneously, and a discharge coefficient of 1.0 is used for the entire analysis. Bernoulli's equation was used for the subcooled portion of the transient while Moody's correlation was used in the two phase portion.
3.
No offaite power is available.
O 6.3-15 Am. 51 (2-4-77)
%d
,,.,-y,...,,---,,*.w
,-----.w---.,,,,g-,,
,i-_,,,__
m..~,e-%--,,-.,.,m--.
--,-w,---.
-.----ew..s-,,y.w-
..w---A-.
o Pages 6.3-17 a:i 6.3-18 deleted vith A=e-e ent 39 e
O.
O O
6.3.3 11 Effects of ECCS Ceeration on the Core The temperature transient in the core can produce significantly higher than ner=al temperatures in components other than fuel rods. Therefore, a possi-bility of eutectic formation between dissimilar core materials exists. Con-sidering the general area of eutectic formation in the entire core and reactor vessel internals, the following dissimilar metals are present, with major elements being in the approximate proportions shown:
Tree 30h Stainless Steel 19% chromium 10% nickel Remainder iron Control Rod Poison Material 80% silver 15% indium 55 cadmium Zirealoy b 98% zirconium 1-3/k% tin O
4 Inconel 53% nickel 19% chromium 35 molybdenum 5% columbium-tantalum 15 titanium 0 55 aluminum Remainder iron All these alloys have relatively high melting points (greater than 2700F) ex-capt for the silver-indium-cadmium alloy, whose melting point is about ikTOF.-
The binary phase diagram indicates that zirconium in the range of 75 to 80% has a eutectic point with either iron, nickel, or chromium at temperatures of approximately 1710, 1760, and 23TOF, respectively. If these dissimilar metals are in contact and if those eutectic points are reached, then the me-terials could theoretically melt even though the temperature is below the melting point of either material taken singly.
One point of such dissimilar metal contact is between zircaloy-clad fuel rods and Inconel-718 spacer grids. The analysis of the loss-of-coolant accident indicated that some of the cladding vill exceed the zirconium-iron and the sirconium-nickel eutectic points. Since the spacers are located at 21-inch intervals along the assembly and each grid has a very small contact area,
'D only a fraction of the hottest fuel rods would be in contact with Inconel-718 spacer grids.
6.3-19
a-
.. n 6.3 3 13 Lag Times The core flooding tanks are self-actuating and vill discharge directly into the reactor vessel when the reactor coolant system pressure falls belov 600 psig. In the analysis of the loss-of-coolant accidents, the high-pressure and low-pressure injection was delayed 25 seconds after receipt of the actus-tion signal.
The testirs procedure in 6.3.h specifies that the valves vill be in their cca-manded (open) position in 115 seconds and the pumps (HPI and LPI) will be delivering their rated flow within 25 seconds. These values agree with the numbers used in the analysis.
6.3 3.lk Thermal Shock Considerations Thermal sleeves are installed where required to limit the themal stresses developed because of rapid changes in fluid temperatures. They are provided in the four high-pressure injection nozzles on the reactor inlet pipes.
Babcock & Wilcox has evaluated the capability of its 177 fuel assembly pres-surized water reactor vessel to withstand thermal shock caused by actuation of the emergency core cooling system following a loss-of-coolant accident without imparing the ability of the vessel to hold water. This analysis is presented in topical report BAW-10018. " Analysis of the Structureal Integrity of a Reactor Vessel Subjected to Thermal Shock".
O Based on the results of the investigation, it is concluded that the reactor vessel vill not lose its in'egrity due to crack propogation resulting from thermal shock caused by actuation of the ECCS following a IOCA even if this transient occurs at the end of 40 years of irradiation and the vessel wall.
contains a flav of critical size.
6.3 3.15 Limits on System Parameters To be consistent with single active failure critiera, the safety analysis pro-vided earlier in this section was conducted assuming the operation of two core flooding tanks, one high-pressure injection train, and one low-pressure injec-tion train. The Technical Specifications require that two core flooding tanks, l
two high-pressure injection trains and two low-pressure injection trains be operable except for short periods when one high-pressure injection system and/or one low-pressure injection train may be unavailable due to maintenance.
Miinimum volumes of the BWST'and CFT are given in 6.3 2.6.
The minimum boron concentration for the water in the BWST and CFT is 2270 pga.
e^
%./
6.3-21
r b
The test of the low pressure injection system is performed when the RC system pressure is at 100 psig. The equipment which would be called upon to operate in the event of an ESF actuating accident is tested. The ESF command is ap-plied to the DE suction valves from the BWST (command open), the M removal pumps (command to start) and the M (low pressure) injection valves (command to open). Each of there devices is considered to have operated satisfactorily vhen it obeys the command as noted. The test is considered to be acceptable when the devices obey their respective commands within the specified time in-
)
terval. Se valves are to be in their commanded position within 11.5 seconds after their respective commands. Each decay heat removal pump (LPI Pump) is to be delivering 3000 gym of borated water from the BWST to the RC system i
within 25 seconds after receipt of the IBF command to " start".
The once per fuel cycle testing frequency of the systems related to emergency core cooling is based upon the shutdown for refueling frequency. The annual test frequency is considered to be satisfactory. The test is considered to give a demonstration of emergency equipment readiness. Refer to Technical Specification 16.4.5. The individual active components within the emergency core cooling systems are tested no less frequently than quarterly (13 weeks) to verify that* the component assumes its ESF commanded status on ESF ce===nA.
S e method of conducting the test is actuating the component through its par-ticular ESF signal. Se device is considered to have " operated acceptably when it goes to,its ESF status within its specified time period (if specified).
The individual components which are tested are shown on Table 6.3-8.
i I
e P.
V 6.3-23
-o TABLE 6.3-1.
ECCS COMPONENTS DATA Core Flooding Tank Number 2
Design pressure, psig 700 Design te=perature, F 300 Operating pressure, psig 600 operating te=perature, F 110 Total volume, ft3 1410 3
Nor=al vater volume, ft loko Minimum boron concentration in water, ppm 2270 Materials of construction CS, SS clad HPI Pumes Number 3
O T/pe Horizontal, multi-stage centrifugal Required NPSh (design), ft 26 Available NPSH, ft 29 Pump material SS, vetted parts ~
Design te=perature/ pressure, F/psig 200/3000 Capacity for HPI requirements, gym /ft 500/2600 LPI/ Decay Heat Removal Pump _
Number 2
Type Single-stage, centrifugal Required NPSH(design), ft.
75 Pump material SS 1
Design temperature / pressure, F/psig 350/520 I
l I m, Max operating temperature / pressure, F/psig 300/h95 d
Capacity for LPI requirements, gym /ft 3000/350 K 1 me Am. 31 (8-15-75)
, L#...,..
s e
O TABLE 6.3-2.
EMERGENCY CORE COOLING SYSTEM Code elassification Component Decay heat removal coolers ASME Section III, Class C (tubeside),
(DE-C1 A and B)
ASME Section VIII, (she11 side), 1968 with Addenda through vinter 1968.
Decay heat removal pumps ASME Pump and Yalve Code for Nuclear Service, (DH-P1 A and B) 1968.
Borated water storage tank AWA-D100-67 except velding procedures.
(DH-TI)
Welder qualifications and joint design is accomplished with ASME Section VIII and II.
A11ovable stress values per Section VIII.
l USAS B31.T dated 1968. (Inspection)
DH injection valves (DE-V4 A and B)
USAS B16.5 dated 1961 (valve design)
Valves between RB sump and USAS B16.5 dated 1961. (valve design)
DE pumps (DE-V6 A and B)
USAS B31.7 dated 1968 (Inspection)
Valves between BWST and USAS 331.7 dated 1968. (Inspection)
DH pump (DE-V5 A and B)
USAS 316.5 dated 1961 (valve design)
USAS 331.7 dated 1968 (inspection),).
i Valves between DE pumps USAS B16.5 dated 1961 (valve design f
and MU pumps (DE-VT A and B) l ASME Section III, Class C, 1968 with Addends Letdown storage tank (makeup tank) (MU-TI) through summer 1968.
ASME Pump and Valve Code for Nuclear Service.
Makeup pumps (MU-P1 A, B, C) 1968.
USAS B31.7 dated 1968 (inspection),).
Outlet valve to makeup USAS B16.5 dated 1961 (valve design tank (MU-V12)
USAS 331.7 dated 1968 (inspection),
IP injection valves USAS B16.5 dated 1961 (valve design).
(MU-Y16 A, B, C, D)
ASME Section III, Class C,1965 with Addendust Core flood tank (CF-T1 A and B) through summer 1967 l
USAS B31.7 dated 1968 and MSS-SP66.
Core flood tank outlet valves (CF-V1 A and B)
ASME Section III, 1971, with Addenda through f
Core flood check valves (CFb, 5 A and B) susmer 1971.
l r-~
lL 6.3-27 l
.1...
~
e TA3I2 6.3-3 (Coct'd)
Design Design temperature,
- pressure, Cocreonent F
vsir Piping from DH cooler outlet to upstream of the reactor building isolation valves 250/300 520/k95 Piping from upstream of the reactor building isolation valves to upstream of the check valves in the core flooding lines 300 2500 Piping from upstress of the check valves in the core flooding lines to the reactor vessel 650 2500 Piping from upstream of the reactor building sump isolation valve to upstream of the valves at the pump inlet 300/250 200/225 O
e e
e e
l I
l,- >
1 %,)
6.3-29
TABLE 6.3-5
^
l RELIEF VALVES CAPACITY AND SETTING IN THE ECCS Capacity (10% accumulation) l Description and settings Function Core flooding tank
+375 scfm at 700 psig To protect CF t:.nk from maxi-relief valve (vent to RB atmosphere) num till rate from makeup pumps.
BWST vacuum breaker 2100 SCFM st 1 os/in2 To protect tank from vacuurz (vent to atmosphere) at navimum draining rate and protect against external pressure during tornado.
BWST relief valve 700 SCFM st 2 5 psig To protect tank from over-(vent to atmosphere) pressurization at marimum fluid volume fill rate.
LPI (DE) pump inlet 4 0 gym at 200 psig To protect against overpres-line relief valve (vent to floor drain) surization due to ambient temperature change and leak-l age from normal DH flow path through closed valves during normal DH system operation.
O LPI (DE) pump dis-4 0 gym at h95 psig To protect against leakage charge line relief (vent to floor drain) from RC system through nor-valve mally closed DH injection j
valves during normal RC sys-tem operation and protect against overpressurization due to ambient temperature change.
Makeup tank relief
%1ho scfm at 100 psig To protect MU tank from maxi-valve (vent to vaste gas zum till rate.
filter) l l
\\
t U s.3-31
4 Table 6.3-6 (Cont'd)
'O Com;enent Malfunction Co:nments C.
Core Flooding System 1.
Isolation valve Closes during If the vr.1ve cannot be manually in discharge line, normal opera-opened, de reactor must be shut tion.
down or operations limited as specified in Technical Specifi-cations.
2.
Tank relief valve.
Opens during Loss of nitrogen pressure and normal opera-consequent loss of ability of tion.
tank to perform. Reactor must be shut down or operations ad-justed to Technical Specifica-tion limits until the relief valve is repaired.
3.
Check valves in Excessive leak It is extremely unlikely that discharge line.
detected during both check valves would permit normal reactor excessive leakage. Leakage operation.
would be indicated by CFT pres-sure and level changes. If leakage becomes progressively worse or is unacceptably high, reactor must he' shut down while O
the check valves are repaired.
i e
4 0
?
V 6.3-33
.L
i TABLE 6.3-8.
A PROCEDURE AND REQUIREMENT FOR ACCEPTABILITY yrecuency Component Procedure HPI pumps Start idle pump via E1 signal, check status lights and developed head.
Quarterly HPI valves-Open HPI valve via ES' signal and verify position via position indicating lights.
Quarterly /
HPI suct valves Open suction valve via ES' signal and verify from makeup tank position via position indication lights.
Quarterly Decay heat removal Start pump via E!r signal, check status pumps (LPI pumps) lights and developed head.
Quarterly Decay heat removal Open valve via ES' signal and verify valves (LPI valves) position indicating lights.
Quarterly Decay heat removal Valve is normally open. Command valve is pump suction valves closed via remote manual switch. When (from BWST) open valve via ES. signal and verify posi-tion via position indicating lights.
Quarterly
'O Decay heat removal Open valve hy remote manual signal
%/
pump suction valves (noes). Verify position via position (from sump) indicating lights.
Quarterly l
l
,D 6.3-35
a c
. a
- r. e.,
m am
.e, pos u&
t u
W
- A uSe
- m. 5 v
1n 3
~
3 n.
6A 0
(
e )a.
2.
~.
'y
=
sfr n M m__
4 I
oI 6.) ln
~
I I
nISi I
o$
3$_
=
3
=
1
=
81
=
SE.
ns
- ._:==..-
=
N
^y
- . :: : = =..
~..
- 3' ',,: ; ;,.,
p
.7
@~.. n ai g ya s
F
.q.
+
. Ii
{
.u-
' x-p.l
+
7
=*
f
+
S m.W wn 9
1 E
Y n,
}
u-4 y
G o e.
,,7
, g
- .p :
-..- e.,
D" t
[
l
_9 J
%e 8p
'._',f t.s he 8
r a,
I.
n 3oa
.in rn -
p9 4:-
4
=
s s
s w
z.
r.
/
-f f,
P
.I pnw4
( 4 g
o.
q y
~
~
q
.. j
= ;
wL gO
~_ ' ; ~_
}
{
.,t g ~g@
3 h
8'
~g+
Q m g
M4y
.t, g_s yt g
g g
b g
=
)
5 r
r 9tt w.$-
g b
ti
~
.i Ii1 c
jl
(
~
[
Jps a
~
9 e 63 q
g t
g f
n g.
_]9rs
,:,M~r rA 4
7 4
g g
v n.
h s
g
-=
g4 5
p i
.e,~ l
. 1P g
%q 4
~=
e 4
~2 il'i l
~
Total Developed Head, ft E
E E
3 3
3 3
3 a=
9 I
E l
=
/
I g
f,'
1 3
l
=
3 l
n.
e.
1 f
w-4 I
E 1
E
[
l 1
1 E
7
/
I
,E I
I I
I E
2 l
3 5
=
l.
w-MAKEUP PUMP CHARACTERISTICS THREE MILE ISLAND NUCLEAR STATION UNIT 2 N kl h FIGURE 6.3-2
^
~~
AM. 38 (2-20-76)
,.l
- . e I
m 4
N i
W
~
E
\\
3 l
/
N I
l g
w I
in R'
> C'-
f
=
5 l-WSil I
8 I
500 tesa 1588 2000 2588 3008 3588 4000 Capacity, 8ps DECAY HEAT REMOYAL PUMP CHARACTERISTICS THREE MILE ISI.AND NUCLEAR STATION UNIT 2 FIGURE 6.3-3 s*y*
O 60 NOTES: I-THI6 CURVE SUPER 55055 FIG. 20 IN StW LIMITS 4 PRECAUTIONS FOR TMi UMIT 2, DATED 5-31-77 2-INSTRUMENT ERRORS ARE NOT o.
IMC.LUDED.
l.0 -
3*'
~
o nb eE 40 -
O to t
g bJe 30-Q.
20-Io -
0 Q
20 M
leo 00 ICO IMOICATED LEVEL IM MU-T-I i
MAXEUP TANK - LEVEL VS PRES 3URE ggggg THREE MILE ISLAND NUCLEAR STATION LHIT 2 l
l l @ P1GURE 6.3-34 AM. 45 (3-11-71)
g a.r ** 8 *.
- O e
t FIC M S 6.3-4 TEROUCE 6.3-54 DELETED WITH AHENDHENT 41 FIGURES 6.3-55 TEROUGE 6.3-62 DELETED WITH AMENDHENT 51 l
O Am. 51 (2-4-77)
- e
- - - - -