ML20005B968
| ML20005B968 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 09/09/1981 |
| From: | FRANKLIN INSTITUTE |
| To: | |
| Shared Package | |
| ML20005B965 | List: |
| References | |
| TASK-05-10.B, TASK-5-10.B, TASK-RR TER-C5257-311, NUDOCS 8109160211 | |
| Download: ML20005B968 (89) | |
Text
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LNCLOSURE 2 fEP Review of Safe Shutdown Systems for the Haddam Neck Plant l
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4 TER-CS257-311 CONTENTS Section Page 1
INTRODUCTION.
B-1 2
DISCUSSION B-5 2.1 Normal Plant Shutdown and Cooldown.
B-5 2.2 Shutdown and Cooldown with Loss of Offsite Power B-8 3
CONFORMANCE WITH BRANCH TECHNICAL POSITION' 5-1 FUNCTIONAL REQUIREMENTS B-9 0-10 3.1 Background.
3.2 Functional Requirements.
B-23 f.
3.3 Safe Shutdown Instrumentation B-58 4
SPECIFIC RESIDUAL HEAT REMOVAL AND OTHER REQUIREMENTS OF BRANCH TECHNICAL POSITION 5-1.
B-61 4.1 Residual'Eeat Removal System Isolation Requirements.
B-61 B-64 4.2 Pressure Relief Requirements
'B-69 4.3 Pump Protection Requirements B-70 4.4 Test Requiremer..s B-71 4.5 Operational Procedures.
B-72 4.6 Auxiliary Feedwater Supply.
5 RESOLUTION OF SYSTEMATIC EVALUATION PROGRAM % PICS B-73 5.1 Topic V-10.B RER System Reliability B-73 5.2 Topic V-ll.A Requirements for Isolation of B-73 High and Low Pressure Systems 5.3 Topic V-ll.B RHR Interlock Reqcirements B-74 5.4 Topic VII-3 Systems Required for Safe Shutdown.
B-75 B-77 5.5 Topic X Auxiliary Feed System.
B-79 6
REFERENCES
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TER-CS257-311 1.
INTRODUCTION The Systematic Evaluation Program (SEP) review of the " safe shutdown" subject encompassed all or parts of the following SEP topics, which are among those identified in the Nuclear Reactor Pegulation document entitled "Peport on the Systematic Evaluation of Operating racilities" [1):
1.
Besidual Heat Pemoval (RER) System Peliability (Topic V-10.B) 2.
Pequirements for Isolation of High and Low Pressure Systems (Topic V-ll. A) 3.
RHR Interlock Requirements (Topic V-ll.B) 4.
Systems Pequired for Safe SSutdown (Tcpic VII-3) l 5.
Station Service and Cooling Water Systems (Topic IX-3) 6.
Auxiliary Fe-dwater System (Topic X).
The review was primarily performed during an onsite visit by a team of SEP per sonnel. This ensite effort, which was performed from July 11 to 13,1978, af ferded the team the opportunity to obtain current information and to examine the applicable equipment and procedures.
The review included specific system, equipment, and procedural requirements for remaining in a hot shutdown condition (reactor shutdown in acccedance with technical specifications, temperature between 200*F and 350*F) l and for proceeding to a cold shutdown condition (temperature less than 200*F).
The review of the transition from normal operation to hot shutdown l
considered the requirement that the capability exist to perform this evolution from outside the control room. The review we s augmented as necessary to assure resolutien of the applicaole topics, except as noted below:
':tpic V-ll. A (Pequ?.rements for Isolation of High and Iow Pressure Systems) was examined only for application to the RER system. Other high pressure / low pressure interfaces were not investigated.
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TER-CS257-311 systems that are vital to the performance of safe shutdown components.
The criteria against which tne safe shutdown systems and components were compared in this review are taken from the Standard Review Plan (SRP) 5.4.7,
" Residual Heat Removal (RER) System"; Branch Technical Position RSB 5-1,.
" Design Requirements of the Residual Heat Removal Systems"; and Regulatory cuide 1.139, "Guid nce for Besidual Eeat Removal." These documents represent current staff criteria for the review of applications for operating licenses.
This comparison of existing syst*ms against current licensing criteria led' naturally to at least a partial comparison of design criteria, which will be per tinent to SEP Topic III-1, " Classification of Structurcs, Components and Systems.(Seismic and Quality)."
As noted above, consideration of the, six topics did not take into account possible interactions with other tepics, systems, and cm ponents not directly related to safe shutdown. For example, Topics II-3.B (Flooding Potential and Protection Requirements), II-3C (Saf ety-Related Water Supply), III-4.C (Internally Generated Missiles), III-5.A (Ef fects of Pipe B:eak on Structures, Systems, and Co=ponents Inside Containment), III-6 (Seism.w Design Considerations), III-10.A (Thercal Crierload Protection 'for MotErs of Motor-Cperated Valves), III-ll (Component Integrity), III-12 (Environmental Qualification of Saf ety-Related Equipment), and V-1 (Compliance with Codes and Standards) are among other topics which might be affected by the results of the saf e shutdown review or which might have i safety impact upon the systems l
wnich were reviewed. These effects will be determined by later review. Also, this review did not cover, in any significant detail, the reactor protection system or the electrical power distribution system, both of which will be reviewed later in the SEP.
" The major factor considered in assessing the safety margin of each SEP f acility is the ability to adequately protect against postulated design basis eve nts (DBEs). The SEP topics provide a major input to the DBE review, from the standpoint of assessing both the probability and consequences of certain A
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TER-C5257-311 events. As examples, the safe shuticwn topics pertinent to the listed DBEs (2) are listed in Table 1 (the extent of applicability will be determined during the DBE review for Haddam Neck).
Although limited in scope as r:cted above, the safe shutdown tcpic review contributes significantly to the assessment of the existing safety margins at the Haddam Neck plant.
Picine System Passive Failures The NRC staff normally postulates piping system passive failures as (1) accident-initiating events in accordance with staff positions on piping failures inside and outside containment', (2) system leaks during long-term coolant recirculation following a LOCA, and (3) failures resulting frem In this evaluation, certain hazards such as earthquakes and tornado missiles.
piping system passive f ailures have been assumed beyond those normally pcstulated b'y the staf f, e.g., the catpstrophic failure of moderate energy systems. These assu::ptions were made to demonstrate safe shutdewn system redundancy in the event of ecmplete failures of these systems and to f acilitate future SEP reviews of DBEs and other tcpics that will use the s'afe i
snutdown evaluatien as a source of data for the SEP facilities. SRP 5.4.7 and B"'P RSB 5-1 do not require the assu=ptions of piping system passive f ailures.
Credit for Coeratina Procedures For the safe shutdewn evaluation, the staf f may give credit for f acility cperating procedures as alternate means of meeting regulatory guidelines.
l These precedural requirements identified as essential for acceptance of a SEP topic en DBEs will be carried through the review process and considered in the integrated assessment of the facility.
At that time, the staf f will decide which procedures are so important that an aCministrative procedure should be
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established to ensur.e that, in the future, these operating precedures are not 1
l changed without appropriate consideration of their importance to the SEP tcpic evaluation.
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TER-C5257-311 TABLE 1 DESIC : BASIS EVENT IMPACT UPON P DBASILITY SAFE SHUrDOWN 'IOPIC (DBE)
OR CONSEQUENCES OF DBE V-10.B VII (Spectrum of Ioss-of-Consequences C'alant, Accidents)
V-ll.A VII (Defined above)
Probability V-ll.B VII (Defined above)
Probability o.
VII-3 All (Defined as a generic topic)
Consequences IX-3 III (Steam Line Bre&X Inside Consequences Containment)
(Steam Line Break OutsiGe Containment)
IV (Loss of AC Power to Station Cor2*quences f*
Auxil.i ary)
(Ioss of All AC Power)
V (Loss of Forced Coolant Flow)
Probability (Frimary Pump Rotor Seizu *)
(Primary Pump Shaf t Breal.)
VII (Defined above)
Consequences X
II (Ioss of D'ternal Ioad)
Consequences (Turbine Trip)
(Ioss of Condenser Vacuum)
(Steam Pressure Regulator Failure)
(Ioss of Feedwater Flow)
(Feedwater System Pipe Break)
III (Defined above)
Consequences IV (Defined above)
Consequences V
(Defined above)
Consequences VII (Defined above)
Consequences 4
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i TER-C5257-311 2.
DISCUSSION 2.1 t2ermal Plant Shutdown and cooldown A series of controlling peccedures is used to progress from full power to cold shutdown. The first procedure is Normal Operating Procedure (NOP) 2.2-1,
" Changing Plant Load," by which the load is reduced en the turbine generator unit. Throughout the power reduction, the reactor coolant system is borated with the chemical and volume centrol systen (CVCS) as necessary to maintain the proper shutdown marg'in on the control rods. When reactor power is reduced to 27 5-3 0 0 MW, a feedwater pump is taken out of servicer at 230-250 MW,
a condensate pump is taken out of service. The condensate water cources are, in order of preference, the hot well, the demineralized water storage tank (LWST), and the primary water storage tank (FWST). At appecximately 80 MW,
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' the following steps are performed in accordance with NOP 2.3-1,
" Minimum Load to Het Standby":
1.
Phintain T ve $3312*F by maaual adjustment of centrol a
grouc rod position.
2.
Kaintain ster
aratcr at normal cperating level (25-50%)
on narrow rr ng manual feedwater control.
3.
Reset steam dump controller to 880-930 psig.
4.
Transfer buses lA and 1B frem the turbine generator to offsite power. This provides of fsite power to the reacter coolant pumps.
5.
Unload and separate the turbine generater from the system (NOP 2.16-2).
6.
Peduce power to below S0 MW, with manual insertien of control rods until power is at 1 x 10-7 amperes en the Intermediate Range (IR) indicatcrs.
At this point, the plant is in hot stand 5y, and manual adjustments of the controlling group of rods maintains the reactor power level at 1 x 10' at 533 1 2*F.
The pressure of the main coolant ampere s (IR) and the T,y, will be automatically centrolled and maintained at 2000 1 25 psig.
The secondary plant (turbine generator and auxiliaries) is in a het standby condition with the unit en its turning gear, steam seals on, normal B-5 J... Franklin Research Center 4 e-et w.~e w.r.
TER-CS 257-311 vacuum, steam generator levels manually controlled, and a minimum number of auxiliaries in operation. An alternate source of auxiliary electrical power is available.
Normal Operating Procedure 2.3-2, "Eeactor Shutdown," describes the next stepr. in the plant shutdowns inserting the control rods to the 10-step limit and then opening the trip breakers. It is also possib? e to shut down the reacter by automatic action of the reactor trip system or by manual reactor trip by the operator. These actions and the necessary manual procedures for the, operator are described in Dnergency Operating Procedu.2 (EOP) 3.1-1,
" Emergency Shutdown."
The reactor is now maintained in the hot shutdown condition with the reactor suberitical in accordance with NOP 2.3-3, " Operation at Hot Standby-heactor Shutdown." The reactor coolant syst.4 tamperature is maintained at the approximately normal no-load value of 533*F with reactor I*
residual heat and a minim'as number of rcactor coolant pumps operating.
Steam generator pressure is mai.'tained by the steam dump system or by venting to atsosphere using the atmospheric dump valve (ADV).
Steam generator water level is maintained by use of one a miliary feed pump or a mair feed pump under manual contrel. Makeup feedwater is available from the condenser or DWST.
The Technical Specifications require a minimum of 50,000 gallo.is of water in the DWST and a minimum cf 80,000 galle".s of water in the PWST. Water may be transferred from the PWST to the DWST at the rate of 200 gpm with transfer l
pumps. A third source of water is available from the racycle primary water storaga tank (RPWST). There are.no Technical Specifications requirements for l
the amount of water in the RPWST, but the Licensee normally maintains 95,000 to 100,000 gallons in this tank.
If heat is removed t! rough the atmospheric vents, the source of water for the auxiliary f eedwater pumps is the DWST.
If the steam dumps a:e used, water in the hotwell is the source and the excess water is directed through the condensate pumps to the DWST and then back to the auxiliary feedwater pumps.
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TER-CS257-311 The next proceducts, NOP 3.3-4, " Hot Standby to Cold Shutdown," describes a method of cooling cown and depressurizing the reactor frcm the hot shutdo:7 to e refueling condition in which the reactor ecolant terperature is maintained below 150*F.
Prior to cooldown, it is determined that the boric acid and primary makeup water systems have enough capacity to cocpensart for the reactor coolant shrAnkage. Cooldown then proceeds in the following steps:
1.
Borate the reactor coolant system (RCS) to the cold shutdown or refueling boren concentration.
5 2.
Switch of f all pressurizer heaters.
3.
Begin cooldown cf the RCS by venting steam to the steam dump or the atmosphere (ADV).
8 4.
Begin cooldown of the pressurizer u.d depressurization of the pressurizer and RCS with the pressurizer sprays.
5.
Wh'en pressurizer temperature reacbes approxit..stely 450*F and pressure is appecxiutely 600 psig, collapse the steam bubble and fill the presprizer.
6.
When RCS terperature reaches 340*F and pressurizar temperature is less than 43C'F, reduce RCS pressure to 350 psig and place the low tenperature overpressurization relief valves in service (relief setpoint is 380 psig).
}
7.
Place the RER system in operation WP.en RCS pressure has been reduced to 300 psig and RCS temperature is 350*F.
j 8.
When steam generator pressure reaches 300 psig, take the auxiliary feed pump out of service and supply feedwater to the stc an generators with the condensate pumps.
9.
Centinue cooldewn with the RER system antil the RCS meets the cold shutdown requirements. The cooldown procedure calls fee intermittent cperatien of a sing?.e reactor coolant pump which requires repetitive operation of the RHR isolatf en valves during cooldcWn.
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TER-C5257-311 The norma 1 operating pressures of the system used for shutdown cooling are 150 psig plus the pump suction for the RER loop, 82 psig for the component cooling loop, and 55 to 70 psig for the service water system. Therefore, the flow of impurities should be away from the reactor coolant system.
1.2 Shutdown _ and Cooldown with Loss of Of f site Power i,
The shutdown following a loss of offsite power is achieved with EOP 3.1-9, " Total Less of AC."
A loss of ac power ressits in the loss of circulating water pumps, the muin feedwater pumps, and the condensate pumps:
the reactor coolant pumps remain in service for approximately 1 minute (if a
' turbine was operating when of fsite power was lost) and then undergo a gradual coastdown prolonged by the inertia of the fly wheels on the pumps. The pump coastdown time is estimated to be 3-4 minutes with no pumps running (i.e.. no back flow), according to the March 1968 repor t. " Natural Circulation Test of Reactor Coolant System."
The operator is directed tG verify that the steam dump system operates, t' hat the diese1. generators start, and that the neceurary electrical switching is completed to remove connections to the of fsite power lines and to connect the diesel generators to 4160-V ac buses. An auxiliary feedwater pump is started to restore the steam generator level. The operator is also directed to start a component cooling water pump.
A service water pump will automatically start when diesel generator power is available. A charging pump is available to maintain primary ecolant inventory.
Since the procedure requires reaching only a hot shutilaari condition, the operatc,r is instructed to energize the pressurizer heatars. The procedure specifies alterncte sources of water for the auxiliary feedwater pimps if the primary supply is exhausted (Reference 16).
The site experienced loss of offsite power on April 27, 1968 and July 15,
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1969. when the unit was on line, and on June 26, 1976, when it was being re-fueled.
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I TER-CS 257-311 3.
CONFCPy.ANCE WITH BFA!CH TECHNICAL POSITICN 5-1 FUNCTIONAL REQUIRDENTS The current NRC criteria used in the evaluation of the design of.the systems required to achieve cold shutdown for a new facility are listed in Standard Review Plan (SRP) 5.4.7, Branch Technical Pesition RSB S-1, and Regulatory Guide 1.139.
The following paragraphs give a point-by-point comparison of E anch Technicel F?tition RSB S-1 with the shutdown systems at the Haddam Neck plant.
Branch Technical Pesition RSB S-1 Functional Recuirements "The system (s) which can be used to take the reacter from normal operation conditions to cold shutdown shell satisfy the functional requirements listed below.
I 1.
The design shall be such that the reactor can be taken from normal-operating conditions to cold shutdown
- using only saf ety-grade systems. These systems shall satisfy General Design Criteria 1 through S.
2.
The system (s) shall have suitable redundancy in components and features, and suitable interconnect',ons, leak detection, and isolation capabilities to assure that for onsite electrical pcwer system operatien (assuming of fsite power is not available) and for of fsite electrical power syLtem operation (assuming ensite power is' not available) the system function can be accomplished assuming a single failur'e.
3.
- te rystem(s) shall be capable of being cperated from the centrol roem with either only onsite or only cf fsite power available with an assumed single failure. In demonstrating that the system can perform its function tssuming a single failure, limited operator action cutside of the centrol reem would be censidered acceptable if suitably justified.
4.
The system (s) shall be capable of bringing the er. actor to a cold shutdown condition *, with only of f site or or.s;t~ r power available, within a reasonable period of time following statdown, assuming the most limiting single failure."
"*Pree. esses involved in cocidown are heat removal, depressurization, flow circulation, and reactivity control. The cold snutdcwn condition, as described in the Standard Technical Opecifications, ref ers to a suberitical reactor with 4 reactor coolart temperature no greater than 200*F."
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i TER-C5257-311 Compliance of the Haddam Neck safe shutdown systems with these criteria l
is discussed below.
3.1 Backcround The BTP 5-1 requirerents are stated with respect to plant shutdown and j
ecoldown with only of fsite or only onsite power available. The staff i
evaluated the plant's ability to conduct a shutdown with only offsite power I
available and determined that the "only onsite power available" case is more limiting. The plant electrical system is sufficiently versatile.to allow energizing of all necessary equipment from only of fsite power. Therefore, the
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staf f concentrated its evaluation of safe shutdown systems on those needed for shutdown following a loss of offsite power.
A " safety-grade" system is defined in the NUREG-0138 [3] diseassion of i
ifrue No.1 as one which is designed to seismic Category I (Regulatory Guide 1.29) and Quality Group C or better (Regulatory Guide 1.26) specifications and is operated by electrical instruments and controls that meet Institute of Electrical and Electronics Engineers Criteria for Nuclear Power Plant Systems (IEEE Std 279-1971). The Haddam Neck plant received its Provitional Operating License on June 30, 1967 and its Full Term Operating License on December 27, 1974, so that the plant was designed and constructed prior to the issuance of Regulatory Guides 1.26 and 1.29 (as Fafety Guides 26 and 29) on March 23,,1972 and June 7,1972, respectively. Also, proposed IEEE Std 279, dated August 30, 1968, was not used in tbc design of Haddam Neck instrumentation and control systems.
In addition, the Haddam neck plant was built and licensed prior to the issuance of the proposed Genertl Design Criteria on July 11, 1967.
Therefore, for this evaluition, systems which should be " safety grade" are the systems identified in the minimum list of safe shutdown systems.
General Design Criterion (GDC) 1 requires that systems important to safety be designed, fabricated, erected, and tested to quality standards, that a Quality Assurance (QA) program be implemented to assure these systems will perform their safety functions, and that appropriate records of design, fabrication, erection, and testing are kept.
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i TER-CS257-311 Regulatory Guide (RG) 1.26 provides the currient NRC Oriteria for quality group classification of safety-related systems. 71though RG 1.26 was not ir, effect when Paddam Eeck wsc constructed, the Licensee has since classified the systems in accordance with NUFIG-0138 [3]. Also, ween though the safety-related systems were not designed, fabricated, erected, and tested using RG 1.26, the maintstnance and repair of the classified systems is currently conducted in accordance with this guide.
In the Facility Description and Safety Analysis Report for the Laddam Neck plant (4), the Licensee has identified maximum seismic ground accelerations which were used in the design of structures, systems, and comynents important for nuclear safetyi Several systemr, including the reacter coolant, high and low pressure safety injection, RHR, and CVCS, have been designed for a 0.17 g maximum ground acceleration, and the remaining structures and elements of the plant are considered capable of withstanding the seismic forces corresponding to a ground acceleration of at least 0.03' g.
No structures or equipment are classified as seismic Category I per RG 1.29.
Therefore, ground acceleration levels rather than seismic design classifica-tions,have been assigned to safe shutdown system.
At the time the Haddam Neck plant was originally licensed, the NRC (then the AEC) QA criteria had not been developed. However, the QA program for plant operations wa's reviewed by the staf f and found to conform with 10CFR50, Appendir 3 [5). Appropriate records concerning design, fabricaticn, erection, and testing of equipment i.mportant to safety are maintained by the Licensee in accordt.nce with the QA program and the plant Technical Specifications.
GDC 2 states that the structures and equipment important to safety shall be designed to withstand the effects of natural phencmena without losing their ability to perform their safety functions. Natural phenomena considered are hurricanes, ternadoes, floods, tsunami, stiches, and earthquakes.
Measures were taken in the design of the plant to protect against floods a:id earthquakes. During the Full Term Operating License review, the staf f agreed with the Licensee's conclusions [6] that the ability of the structures D
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TER-CS257-311 I
and equipment important to safety to withstand the effects of tornadoes, floods, earthquakes, winds, ice, and other local disturbances was acceptable.
The effects of tornadoes will be reevaluated during the course of the SEP-in Topics II-2.A, " Severe Weather Phenomena"; III-2, " Wind and Tornado-Loadings"; and III-4.A, " Tornado Missiles." The effects of flood will be
- eassesse'd in the SEP review under Topics II-3.B, " Flooding Potential and Protection Requirements," and III-3, "Eydrodyarnld L.nads."
Also within the SEP review, the potential for and consequerces of a seismic event will be reascessed under several review topics.
GDC 3 requires that structures, systems, and components important to safety be designed and located to minimize the effects of fires and explosions.
A Haddam Neck fire protection Safety Evaluation Report was completed on October 3, 1976. The Haddam Neck fire protection reevaluation resulting from the Browns Ferry fire is currently underway in the NRC Division of Operating 9
Reactors. The results of this reevaluation will be integrated into the SEP assessment of the plant.
GDC 4 requires that equipment important to safety be designed to w* thstand the effects of envirencental conditions for normal operation, I
maintenance, testing, and pcstulated accidents. Also, the equipment should be i
protected against dynamic effects including internal and external missiles, l
pipe whip, and fluid impingement.
l I
Tne SEP will consid:r various aspects of this criterion during a review of Topics III-12, " Environmental Qualification of Safety-Related Lquipment";
III-5.A, "Ef fects of Pipe Dreaks Inside Containment"; III-5.B, " Pipe Breaks l
Outside Containment"; and III-4, " Missiles Generation and Protection."
GDC 5 is not applicable to the Baddam Neck plant because the plant does l
not' shar e any equipment with other facilities.
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In order to accomplish a plant shutdown and cooldown following a loss of of fsite power, certain " tasks" must be performed, such as core decay heat removal, steam generator makeup, and coolant boration. The staff and Licensee developed a " minimum list" of systems necessary to perform these tasks l
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1 TER-C5257-311 considering a less of ac power and the most limiting single failure. The systems were then evaluated with respect (1) their ability to perform these tasks and (2) the ' functional requirements of BTP 5-1.
The minimum list of systems (or components) necessary for safe shutdown follows:
1.
main steam - syster atmospheric dump valve, steem generator vents, and auxiliary vent paths 2.
auxiliary feed system 3.
water sources - demineralized water storage tank (DWST), primary water storage tank (PWST) 4.
residual heat removal (RER) system
)
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component cooling water (CCW) system 6.
service water system 7.
chemical and volume control system (CVCS) and refueling water storage tank (RWST) 9 pressute control and relief system 9.
control air system 10.
emergency power systems (ac and de) for the above equipment 11.
instrumentation for the above equipment.
The staf f's evaluation of each of these systems with respect to the BTP
=~ ~ functional requirements is given in succeeding sections. Subsystems and cr.rponents associated with the respective systems are listed in Table 3.1.-
with quality group classifications and ground acceleracion levels. Tables 3.2-1 through 3.2-9 in the next section list the major safe shutdown components, their locations, an evaluation of each system, and the associated power supplies.
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TABIE-3.3 CIJCSIFICATION OF SIRITDOWN SYSifMS - IIADDAM NECK PLAfff Quality Group 7
Seismic Plant Plant
'i.
Compone nt s/ Subsystems R.G.
1.26 Design R.G.
1.29 Design Remarks
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REAC1DR CONTROL AND NA*
Category I Note 1
- NA = not applicable gh PROTECTION SYSTFM Note 1: All systcms yy and equipment are M}$
' withstanding the considered capalle of seismic forces cor-
!"g responding to a ground acceleration of at 1 cast 0.03 g unless otherwise noted in this table (Ref. FDSAR [4] Section 2.5)
Y MAIN STFJW SYSTFM g
MS atmospheric relief ASME III ASME VIII Category I 0.03 y (IIICV-1201)
Class 2 Piping and va1ves from the ASME III AS A B31.1 Category I 0.03 g steam generator to and Class 2 including MS isolation valves, IIICV, PICV-1206A and
-1206B valves, TV-12?? ara
-1213 drain and trap iso-lation valves, and V-15,
-16, -25, -26, -35, -36,
-4 5, and -46 vent isolation valves Piping fro.a PICV-1206A ASME III AS A B31.1 Category I 0.03 9
- Plant design informa-and -1206B valves to Class 3 tion obtained f rom auxiliary feed pumps FDSAR 14) and CYPPCO including SV-1216A and letter [7]
-1216B valves s
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TABLE 3.1 (Cont.)
Quality Group Seismic Plant Plant E2 Cosnpone nts/Subsys tems R.G.
1.26 Design R.G.
1.29 Design Remarks Pipiro and valves from MS ASME III
?
Category I
?
isolation valves to hogging CLASS 3 g
jets and sirvjle jet air W
ejectors EEjK AUXILT ARY FEED SYSTEM (AFS)
Ia AND WATER SOURCES
-a Turbine-driven pumps (2)
ASME III Category I 0.17 g Class 3 Pipirvj and valves f rom ASME III ASA D31.1 Category I 0.17 g pump discharge to valves Class 3 156-1 through -4 and 102 m
Domineralized, water ASME III ASA B90 Category I 0.17 g storage tank (0WST)
Class 3 PipimJ and valves to ASME III AS A H31.1 Category I 0.17 g suction of AES purps Class 3 from DWST Pipimj and valves from ASME III ASA B31.1 Category I 0.17 g valves 156-1 through -4, Class 2 102, main feed valves MOV-Il through -14 and 135-1 through -4 to steam generators Q
re h
Primary water storage tank ASME III
?
Category I
?
(PWST)
Class 3 g
-J B
PipimJ and valves f rom ASME III
?
Category I
?
y PWST to DWST, including Class 3 P
primary water transfer pump (IMTPs)
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TABI.E 3.1 (Cont.)
J ua1ity Group Seismic Plant Plant
- g, 8'
Components / Subsystems R. G.
1.26 Design R.G.
1.29 Design Remarks 9
F.
((
RESIDUAL llEAT REMOVAL t 5' (RilR) SYSTFM N
IU RIIR pumps (2)
ASME III W spec.
Category I 0.17 g RilR pumps provide h
Class 2 ASME VIII ECCS contairunent
[y ASME IX recirculation Eg
~$
RIIR heat exchangers (2)
(tube side)
ASME III ASME VIII Category I 0.17 g Class 2 Case 1720 N (shell side)
Piping and valves to RIIR ASME III ASA D31.1 Category I 0.17 g Z
pump suction from refueling Class 2 Special water storage tank (PwST),
Nuclear containment sump, and Cases valve 781 t lping and valves from IUIR ASME III ASA B31.1 Category I 0.17 g i
pump discharge (709-A,-D)
Class 2 to the reactor coolant system (RCS) (003) via the RIIR heat exchangers, RitR bypass (FCV602), and recir-culation line to RIIR pumps H
MN Piping and valves from RilR ASME III AS A B31.1 Category I 0.17 g h
to chemical and volume Class 2 g
control system (CVCS) pump suction and RwST (valve y
7 24) ; to core deluge; to y
P charcoal filter spray; to reactor containment spray i
~
TAllLE 3.1 (Con t. )
g
- j..
- ,i Ouality Group Seir.mic 55 Plant Plant Components / Subsystems R.G.
1.26 Design R.G.
1.29 Desion Remarks En
{n COMPONENT COOf.ING SYSTEM sl9 Component cooling pumps (3)
ASME III '
No codes Category I
?
?
exchangers (IlXs) (2)
CLASS 3 Case 1270N Component cooling surge ASME VIII ASME VIII Category I
?
tank Case 1270N Y
All piping and valves ASME III ASA B31.1 Category I
?
U associated with the Class 3 Section 1 component cooling water (CCW) system SERVICE WATER SYSTD1 (SMS)
SWS pu:aps (4)
ASME III Industry Category I 0.17 g Class 3 Standards Piping and valves for ASME III ASA B31.1 Category I 0.17 g containment cooling up Class 2 to and including valves 263 through 270 Piping and valves excluding ASME III ASA B31.1 Category I 0.17 g Tube sides of CCW and those almve and up to and Class 3 spent fuel IIXs are including valves 606, 282, ASME VIII and MOV-1, 2, 3, and 4 l
e
Tant.E 3.1 (Cont.)
g Quality Group Seismic 04" Plant Plant Comporents/ Subsystems R.G.
1.26 Design R.G. 1.29 Design Demarks is s 5' ClifMICAI, AND VOI.tME N
~CONTitOs, SYSTIM FDSAR Table 5.2.1-1
.N
[4]
m fy Charging pumps ASME III ASME VIII, IX Category I 0.17 g Class 2 Ilydraulic j$n Institute STD 4
Piping (loop 1) letdown ASME III ASA~
Category 1 0.17 g valves to and including Class 1 D31.1 letdown isolation valves Regenerative heat ASME III ASME VIII Category I 0.17 g exchangers (3)
Class 1 Cases 1270N
~
y and 1273N w
cn Piping and valves (pump ASME III ASA B31.1 Category I 0.17 g oischarge) from and Class 1 including valves 399 and 296 to RCS Piping and valves from ASME III ASA B31.1 Category I 0.17 g pump discharge to con-Class 2 tai rrnent isolation valves 399 and 296 Piping from pump discharge ASME III ASA B31.1 Category I 0.17 g g
via reactor coolant pumps Class 2 y
and from 'IV-1847 to seal A
water llX y
U Piping loop drain line via ASME III ASA B31.1 Category I 0.17 g 4
cooler to and including Class 1 g
valves RV-1847 and TV-1847
TABLE 3.1 (cont.l '
Quality Group Seismic li' '
Plant Plant ComponentsfSuLsyst ems R.G.
1.26 Design H.G.
1.29 Design Remarks
{ ct s 5' Pipirvj and valves down-ASME III ASA B31.1 Category I 0.17 g fA3 stream ot letdown isola-Class 2 tion valves (FCV-202, 203, and 204) to the VCT includ-
[n irv) ' ICV-ll3A via the reactor l'l coolant filter (includi ng p
3 valve 343A), via relief valves 205 and 252, and via
'11:V-1130 1 A p irvj and valves down-ASME III AS A B31.1 Category I 0.17 g stream of TCV-ll3A via Class 3 demineralizers to valves y
343A, 220, 234, and 235 and g
includitcJ the demineralizer fill and drain valves Volume control tank (VCT)
ASME III ASA H31.1 Cateijory I 0.17 g connecting pipirig and Class 2 valves up to valves 1847 (relief), 332 (relief),
' tCV-l l 2C, 246, and 251 (relief), 324, 255, and 317 Piping and valves from VCT ASME III ASA D313.1 Category I 0.17 g to charging pumps up to Class 2 g
and including valves y
MOV-33A, -23D, and -366, f
320, 369, and to the NWST g
via 372 y,
Drain cooler heat exchanger ASME III ASME VIII Category I 0.17 g (tube side)
Class 1 Case 1270N (shell side)
ASME III ASME VIII Category I 0.17 g Class 2 e.
TAllLE 3.1 (Cont.)
Quality Grotp Seismic
+
c Plant Plant r=>
Components / Subsystems R.G.
1.26 Design R. G.
1.29 Design Remarks Non-regenerative heat ASME III ASME VIII Category I 0.17 9 ff exchanger (tube side)
Class 2 Case 1270N
'E (shell side)
ASME III ASME VIII Category I 0.17 g
{n Class 3 Ik 0;
Seal water heat exchanger ASME III ASME VIII Category I 0.17 g (tube side)
Class 2 Case 1270N (sh:11 side)
ASME III ASME VIII Category I 0.17 g Class 3 Mixed-bed demineralizers ASME III ASME VIII Non-0.17 g Class 3 portions of y
Class 3 Case 1270N Category I CVCS associated with y
detaineralizers are not required for safe shutdown.
Volume control tank ASME III ASME VIII Category I 0.17 g Clans 2 Case 1270N Seal water injection ASME III ASME VIII Category I -
0.17 q filters Class 2 Case 1270N Refueling water T.SME III
?
Category I 0.1 g storage tank (RWST)
Class 2 g
N Piping and valves from ASME III
?
Category I 0.17 g A
RWST to charging system Class 2 g
valve 372, including valve 3
MOV-24 4
H
~
TAllLE 3.1 (Cont.)
6
[-
Ouality Group Seismic Plant Plant Components / Subsystems R.G.
1.26 Design R.G.
1.29 Design Remarks
- e. 3 fy PRESStHtE CONTHOL AND f{
ItELIEF SYSTEM
(;;
g'c.
?
!3 Class 1
. Case 1270N Z
Case 1273N Piping and valves for ASME III ASA B31.1 Category I
?
spray systems to the class 1 (Piping)
RCS and to MOV-298 Piping for surge line ASME III ASA H31.1 Category I
?
y to the'RCS Class 1 (Piping)
U Piping and valtes from ASME III ASA B 31.1 Category I
?
the pressurizer to the Clace 1 (Piping) pressurizer relief tank via the PORVs Piping and valves from ASME III ASA D31.1 Category 1
?
the pressurizer to the Class 1 (Piping) pressurizer relief tank via the pressurizer safety valves Pressurize;r relief tank ASME VIII
?
Category I
?
Q
- o EMEllGENCY POWER SYSTEMS b
M
-4 1
Distribution li*nes.
NA Category I switchgear, control P
lx>ards, and motor control centers
g, yaIY
- m a
zJ Tant,E 3.1 (Cont. )
ta i
2,7
{:":
Quality Group Seismic
$g 7
Plant Plant l
Components /Subsystemn R.G. 1.26 Design R.G.
1.29 Deslgn I:emar ks 4
e l
Diesel generators NA I
Category I 0.17 g
[7]
l Diesel fuel oil, lube ASMP. III ASA B31.1 Category I 0.17 g oil, and starting air Class 3 DC power supply system NA Category I
~
1 tIf 6
'CONTROI, AIR SYSTEM w
Air compressors (2) 7
?
- Air systems ne N ed for safety functions Al.r receivers (2)
(e.g., accumulators and piping to a safety-related valve) should be quality group C and seismic Category I.
Piping and valves from 7
?
receivers to individual components operated by g
the instrument air system ty e
O b
v.
4
~
i w
e
TER-CS257-311 3.2 Functional Recuirements MAIN STEAM SYSTD1 (ANOSPEERIC DUMP VALVE AND CIHF5 VENT PATES)
Task: To remove core decay heat by venting steam from the main steam system to the atmosphere and to provide a source of steam for the auxiliary feedwater turbines.
Discussion Immediately af ter the loss of offsite ac power causes a turbine trip and reactor scram, steam generator pressure will increase due to residual heat from the sudden load reduction, reactor decay heat, and insufficient steam removal via autcmatic vent paths. During the resulting transient conditior.,
l the main steam safety va,1ves (MSSVs) will open to reliave steam generater pressure and prevent the steam generator from exceeding ASME Section VIII design criteria. The re,lieving action also removes residual heat from the RCS. This transient condition will continue until the heat removal rate riser to equal the RCS decay heat input rate. The plant's condition will then be stabilized so that cooldown can begin. Several steam relieving peths are.
avazlable to remove decay heat, including (1) the atmospheric dump valve (ADV), (2) 1-inch steam generator (SGV) vent lines, (3) auxiliary feed pump (AFF) turoines, (4) hogging air ejectors. and (5) main conden3er single jet I
air ejectors (SJAEs).
i The main steam safety valves have not been included in the minimum i
systems or compentnts list since their purpose is to protect the steam generators by mitigating any overpressure condition caused by a design basis event. They will be included in a separate SEP review of design basis topics.
The air-controlled ADV vents steam from any or all of the 24-inch-CD steam lines via a decay heat release header (DHRH). The DERH is pressurized from the 24-inch steam lines via 3-inch-CD lines just upstream of the non-return valve s.
The DHRH is normally pressurized, and supplies the two turbina-driven auxiliary feedwater pumps as well as the ADV.
Thi DHRH is located in the upper level of the steam and feedwater penetration enclosure immediately outside the reactor containment.
- B-23 dJ Franklin Researc.h Center
% o. a wou.
l
TER-CS257-311 The ADV is a 3-inch-OD valve operated by air pressure against a diaphragm. The position of the ADV varies with the air pressure on the actuator and is controlled from the control room. The ADV fails shut on loss of centrol air and cannot be opened manually.
Upstream of each main steam non-return valve, a 1-inch vent with two manually operated isolation valves permits steam system venting during ECS and steam system starsups. These vents relieve directll to the atmosphere in the steam and teedwater penetration enclosure. Thus, these four 1-inch vents provide another path for steam energy removal.
The auxiliary feedwater pumps (AFPs) provide feedwater to the steam generators during the loss of ac power, and are described in the following section. The AFPs are turbine-driven pumps each rated at 430 hp at full flow. These ATP turbines provide ancther means of removing steam generator (and RCS) energy by venting main steam through the turbine exhausting to the ltbosphere. Scoping calculations indicate that one ATP is capable of maintaining steam generator level within 30 seconds af ter reacter shutdown. A relief valve upstream of each AFP turbine opens at 600 psig and relieves at a rate of about 38,000 lbm/h. Operation of the relief valve will effectively prohibit use of the associated ATP turbine. The relief valves may be manually lifted to relieve steam.
Additional heat is removed through two " hogging jets" (initially used to establish condenser vacuum) and two sets of SJAEs used to maintain condenser The hogging jets are single-stage venturi-type air ejectors which use l
vacuum.
main steam supplied through a pressure reducing valve, draw directly from the condenser steam space, and discharge te the atmosphere.
Each SJAE h;d two first-stage and two second-stage no::les. The first-stage nozzles, which use reduced main steam pressure, draw directly from the condenser steha space and discharge to the SJhE.intercondensor which is cooled by main condensate. The second-stage nozzles also use reduced main steam pressure but draw from the intercondenser and discharge to the after-f condenser which is also cooled by main condensate. Condensed steam from the B -2 4 v.h FitNuin ReMwrh Gn er A ow Tw.
- w*.nm W
TER-C5257-311 inter-and af tar:endensers is directed back to the main condenser, and the accumulated non-condensible gases are vented to the atmosphere.
During cooldown following a loss of ac power, the SJAEs and the hoggers are used to remove energy from the steam generators by bleeding steam from the main steam lines. The normal ;uction paths to the main condenser would be isolated. The hoggers and SJAEs are manually controlled and require adjustments as the steam pressure varies.
The following table shows the earliest time following ac power loss and I
scram when each component identified above can by itself remove core decay heat as fast as it is being added to the RCS. The core decay heat curve was taken from ANS 5.1.
l Steam Flow RCS Full Power Corresponding (lbm/h)
Energy Removal Fraction Time After 6
Component at 1000 psig (Btu /h x 10 )
(t)
Scram (min)
ADV 133,000 141.7 2.26 17 4 1-inch vents 162,700 173.2 2.78
%3 AF?
14,800 17.3 0.27 Hoggers 9,300 11.0 0.17 SJAEs 1,000 1.1 0.01 I
AFP RVs 38,000 41.8 0.67 When the ccmpor.4nt energy removal capability equals the decay heat input rate, plant cooldown can commence and intermittent MSSV lifting will stop, I
Redundancy To establish the degree of redundancy provided by the equipment discussed above, the scaff and the Licensee performed scoping calculations to determine the RCS cooldown times (e.g., from 544*F to 350*T) using various combinations of the components listed above. Results of the staff 's calculations are as follows:
A B-25
- AS! FranMn Research Center aN..aenerr.aonm v.
TER-C5257-311 RCS Cooldown Time Components (h) 1 ADV + 4 S/G Vents + AVPs*
16.8 23.6 3 SGVs + AVPs 4 SGVs 61.1 1 ADV (reaches 371'F in 69.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />)
- AVPs include AFP turhines, hogging jets, SJAEs.
The 16.8-hour period represents the shortest cooldown time with all stean-venting paths in use, no delay before start of cooldown, and no I
lim tation of cooldown rate. The 23.6-hour cooldown time is based on the most limiting conditions which maximize the use of available cooling waters the time includes an initial 4-hour delay, the cooldown rate is restricted to the 50*F/h administrative limit, the AD" is inoperative, and one SGV is assumed to have failed. The two remaining times demonstrate that the active components listed cannot achieve cooldown within a reasonable period (36 h) evan witnout t
both En initial delay and a limit on cooldown rate.
The AD7 fails shut and is inoperable when control air fails due to loss of offsite power. This adversely affects the plant by lengthening cooldown tine and increasing the required inventcry of shutdown cooling ~ water.
CYAPCO has stated that the RCS has been cooled down from about 540*F to 350*F in 12 to 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> using (not simultaneously) the ADV, SGVs, two AFT turbines, SJAEs, and the hogging jets. During the cooldown, the RCS flow was provided by one or more PCPs, which adds heat to the RCS.
Based upon the staff 's calculations, the Licensee's experience, and the available water supplies, the staf f concludes that sufficient steam removal paths are available, without the contribution of the ADV and asseming an additional single active 'f ailure, to cool the RCS to the RHP initiation point before the Technical Specifications minimum of 130,000 gallons is expe;.ded.
Location and Operation The staff evaluated the equipment discussed above with respect to its location and its operability during a loss of of fsite ac power. Table 3.2-1 shows the equipment's location, the places from which it may be operated, and its power supply.
4 B-26 Sb Franklin Research Cewi.
nter
- ts, r,.ae
ff_;
TA5!,E 3. 2-1. MAIN STEAM SYST1H i
>.n EY' 5-IQU I PMEPTP IOCATION
,OPER A"' ION POWER SUPPIX 24' Atirospheric dump tJptwr level of the steam Automatic and manual None lg$
vatve and feedwater penetration from control room y{
enclosure is Hogging jets Intermediate level of local manual opt. ition No ejection pcwer needed the turbine building only near generator end of turbine SJA2s Interaiediate level of Incal manual operation
!!o electrical power needed the turbine building only to 0
1-inch steam vents typer level of the steam Incal raanual o'peration No electrical power needed and feedwater penetration only ciclosure AFP turbines (see discussion of'AFP)
(see discussion of AFP)
(see discussion of AFP)
P.FP turbine relief Adjacent to each AFP RVs are self-actuated when No electrical power needed valves (RVs) turbine turbine casing pressu're reaches 600 psig. The control room operator
(:annot cause the RVs to open automatically but can open then manually.
g N
A Y
M
TER-CS 257-311 AUXILI ARY FEEDWATER SYETD1 Task: To provide steam generator makeup inventory whenever the ROS temperature is > 350'F.
Discussion While the ROS temperature is above 350'F, the core decay heat is removed by bleeding steam from the stcam generators using the various components and flowpaths discussed previously. Because the condensate and feedwater pumps are normally podered from of fsite power, which is assumed lost for this evaluation, and because the emergency diesel generators are inadequate to power these components following loss of the station generator and of fsite power, these pumps will not be available.'
l Sines each steam generator contains about 37,500 lbm of feedwater at full
~
power, about 150,000 lbs are immediately available for primary system energy removal by MSS 7 actuation. Staff calculations of the amount of energy 150,000 lbm would remove by vaporization show that this inventory is suf ficient to maintain an acceptable RCS temperature without requiring any steam generator makeup for approximately 50 minutes. Even if there were a 15-second delay betwean the loss of power (less of load and feedwater) and the reactor scram (i.e., the steam generators would be producing steam at 100% capacity without feedwater), the staf f calculation shows that the steam generatcrs can remove the core decay heat for about 40 minutes before boiling dry.
Af ter a less of of fsite ac power, two turbine-driven, centrifugal AFPs draw water through a buried pipe from the DWST and inject it as feedwater into the four steam generators. The pumps discharge into a common header which branchus into two parallel igths to the main steem generator feed system. One path enters the turbine building via a buried pipe and connects to the bypass lines around the four main feed regulating valves. The other line passes through the containment wall and connects to e:ch stsam generator main feed line downstream of the main feed non-return valve.
O B-28 LUU5ranklin Research Center A om a es n. n e. %.
~ ~
j 1
I TER-C5257-311 k
j The AFT discharge header to the turbine building is equipped with a normally open manual valve in the main steam and feedvater penetration j
j enclosure ar.J a check valve in the turbine building. This line branches and connects to the four main feed regulating valve bypass lines upstream of the air-operated main feed regulating valve bypass valves. The bypass vrives fail open at a lose of air pressure.
The AFP discharge header which enters containment froE the main steam and feedwater penetration enclosure is equipped with a motor-operated valve (MOV-35), which is normally controlled from the control room, and a check valve outside containment. Inside containment, the header branches into four i
supply lines, one for each steam generator. Each line has a normally open manual isolation valve and a check valve. MOV-35 can to powered from MCC 97 f (bus 7-6), energized by' emergency diese'l generator 25, or manually operated.by
+
handwheel in case of electrical power failure.
The steam supply fej the AFPs comes from the DHRR, and the piping design' i
allows either AFP to rec'eive steam from any one or all four steam generators.
j From the CHRH, the steaa. supply for each AFP passes through a normally open manual valve and an air-operated control valve which is normally cperated from l
the control room. The air-operate valves fail shut at loss of air pressurei l
but a hsndwheel on each valve enables local manual startup of an AFP.
i Based on the previous discussion of the stwsm generater boil-off rate l
af ter a scram, sufficient time is available for an operator to manually start i
an AF7 if the control air system is inoperable.
Redundancy Each ATP delivers about 450 gpm of auxiliary feedwater to the steam generator s.
The staff calculated thtt this flow is sufficient to centrol and raise the steam generator level about 30 seconds af ter the scram.
The auxiliary feed system can continue to perform its design function in the event of a single active componeng failure.
Ecvever, the. system is susceptible to single passive failures which could disable both AFPs. These a
B-29
.;.r.l Franidin Rese. arch.C. enter A o
.a d. r, a.u...
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n __,m. - -,,,,. _,,
m,_.,,,_,.,.,-,,._..,,,,
._-..-,_,.---..,n,.
,.,n...,,,,,,,<-
TER-CS257-311 include passivo failure of the common feedwater suction line from the DWST, the common feedwater discharge header, or the decay heat removal header providing steam to the AF7 turbines. In addition, passive failure of the condensate supply line from the DWST to several ncn-essential
- systems could divert condensate away from the AFPs since the AF7 suction line connects directly to the non-essential service header from the DWST.
4 The Licensee has described an alternate method for removing core decay heat if the AFS were disabled by one of the above passive failures. The
" feed-and-bleed" method consists of blowing down the ROS to the primary containment via the pressurizer power-operated relief valves and injecting coolant with the safety injection systems. Long-term cooling would be accomplished by either ECCS recirculation or by the RER system. Containment cooling would be provided by the containment recirculation fan coolers.
The feed-and-bleed mode of operation is not a requirement for any PWRs.
Although the potential for successful core cooling by this me6ns is high, the system is not specifically designed for this mode of operation. For example, the pressurizer safety valves have not been designed to relieve a two-phase or liquid discharge, and it has not been determined whether the design of equipment needed to operated in this mode meets appropriate safety criteria.
l Horeover, this form of cooling ultimately pumps significant quantities of primary coolant into the containment, which is undesircble. The bleed method proposed by the Licensee.also requires the usa of electrical equipment which may be exposed to a harsh environrent as a result of this mode of operation.
l The viability of the proposed feed-and-bleed c.ethod at Haddam Neck has not been demonstrated and these systems have not been included in the vinimum l
systems list. Since consideration of passive failures is not a design l
requirement for residual heat removal systems, this method of ecoling is not needed.
Loss of control air directly affects auxiliary feed system operation by l
necessitating manual actuation -of several key valves in order to provide continuous auxiliary feedwater to remove d-cay heat from the reactor. These actions include manual operation of the AFP turbine steam iniGt valves to l
start the AF7 turbines and manual operation of the feedwater regulating bypass l
l B-30 UEEU Frenklin Research Center
- D-*aa er Th rr., i,..
I I
l
i I
s TER-CS257-311 valves in order to supply auxiliary feedwater to the steam generator and to maintair the steam generator level af ter the reactor scram. These actions will require limited operator action outside the control room in addition to any actions necessitated by a single active failure.
Location and Coeration The staff evaluated the equipment discussed above with respect to its locatf.on and its operability during a loss of offsite ac power. Table 3.2-2 shows the equipment's location, the places from which it may be operated, and its power supply.
ETER SOUFCES - DWST and PWST Task: To supply water to the auxiliary feedwatu system for steam generator makeup.
Discussion Both AFPs take suction from the DWST via the 10-inch hotwell makeup and rejection line. This line leaves the bottom of the DWST and branches into the fellewing:
1.
a 6-inch.hetwell rejection line, i.e., flow from the condenser hotwell using the condensate pump (s) 2.
a 6-inch combined AFP suction 3.
a 10-inch hotwell makeup line 4.
a 3-inch water treatment system line.
The DWST has a capacity of 100,000 gallons and Technical Specifications require a minimum of 50,000 gaillons. Following the loss of ac powoc, the DWST can be filled from the following sources:
1.
the primary water storage tank (PWST) via the primary water transfer pumps (Pv?Ps) or gravity drain 2.
the recycle PWST via the recycle FWTP (s) 3.
the water treatment (WT) system, h
B-31
...U Franklin Research Center a > a n. r.a
,h_fh Tant.E 3.2-2. AUXII.IP.RY FEEWATER SYSTF21
' ?Q c u y
EQUI PMEfff IOCATII)N OPERATION POWER SUPPLY fi*
p.".
Auxiliary feedwater lower level of the steam Operable from the control No electr ical power needed
($
pumps (2) and feedwater penetration room with an air controller g[
- m a (north and south ends) which positions the steam G
inlet valves, or valves
[Q positioned manually if air pressure is lost MOV 35 (single lower level of the steam control room and local MCC 7 (Bus '-6) 7 valve which iso-end fredwater penetration manual lates AF;f from area, above south pump entering contain-to ment - 1 path of Afw 0
flow) u Auxiliary FRVs (4)
Feedwater regulating valve Control room (using air Control air, fall open on area of turbine building nignal) loss of air Feedwater MOVs Feedwater regulating valve Control room (manual) 480-V MCC 5 (Bus 5 or 6)
(must be shut) area of turbine building and local manual or J
Feedwater regulating Feedwater regulating valve Control room Control air, fall open on valves (must be shut) area of turbine building loss of air 4
3 I
H N
A u.
9 M
O
i TER-CS 257-311 The PWST has a capacity of 150,000 gallons ard Technical Specifications require a minimum of 80,000 gallons. The DWST is filled from the FWST by the PWTP.
E3ch of the 2 PWTPs can deliver 200 gpm at 180 psig. The PWTPs have not been included in Table 3.1 because suf ficient water flows by gravity drainage from the PWST af ter the CWST Technical Specifications minimum volume has been expended.
Although no technical specifications require the availability and/or operability of the recycle PWST and the WT system, these are additional sources of water for the steam generators. These systems are not included on the minimum systems list because sufficient water is available for cooldown from the Technical Specifications water inventory in the CWST and the PWST.
lPedundancy Staf f calculations shew that a feed flow of about 69.5 gpm is required free gravity drainage when the required Technical Specifications minimum DWST and steam generator inventory have been expended. If the PWTPs are unavailable due to failure of MCC S-1, then gravity drainage from the PWST to the AFP suction must be at least 69.5 gpn, and the AFP NPSH requirement for this flow (about 17 f t) must be satisfied. The Licensee states that the I
required NPSH is available (8).
The staff calculated the maximum length of time the plant can stay at hot shutdown using the initial steam genere6sc sater inventory and the Technical Specifications DWST minimum inventory. These calculations show that approximately 10.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> of water supply are available before DWST makeup (from the PWST) must be initiated.
The staf f also calculated that the total (Technical i;ecifications required) secondary makeup water inventory (130,000 gallons) it enough either to keep the plant.at hot shutdown or to complete a shutdown to the point of RER initiation (350'F) in abou t 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These calculations take no credit for initial steam generator inventcry or for any condensate in the hotwell.
The previously discussed cceponent cocidown times show that three steam generator vents and auxiliary vent paths can cool the RCS to the point of RER
- e B-33
.a) Franklin Research Center n o.w e n. rven.nu.
TER-C5257-311 initiation (350*F) in about 23.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. However, the staff 's calculations are only scoping calculations, and the Licenses has stated that about 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> are requi:ed for RCS coolderen to 350*F.
Although both' AF7s take suction from the hotwell makeup and rejection line, no credit can be given for the hotwell inventory since the hotwell contents cannot be accurately estimated following a loss of ac power.and subsequent pump trip and reactor scram. However, it is highly likely that a significant amount of condensate (approximately 66,000 gallons) would be available to supplement the previmusly mentioned supplies (i.e., steam generator inventory, DWST, PWST, recycle PWST, and the WT system).
Location and Coeration The staff evaluated the equipment discussed above with respect to its location and its operability during a loss of offsite ac power. Table 3.2-3 shows the equipment's location, the places from which it may be operated, and its power supply.
RESIDUAL HZAT REMOVAL SYSTDi Task: Removal of core decay heat and RCS latent heat to cool the systen from 350*F to 140*F.
l Discussien Af ter the RCS temperature has been reduced to approximately 350'F and the t
pressure is less than 300 psig, the RER loop is placed in service and further l
lowers the temperature to 140*F.
The RHR loop then operates continuously to
(
maintain this temperature as long as required by maintenence or refueling activities.
The RER loop consists of two heat exchangers, two pumps, and associated piping, valves, and control instrumentation. During plant shutdown, coolant is withdrawn from the hot leg of Loop 1 through a single letdown line, pumped through the tube side of the RHR heat exchar.get s, and returned to the cold leg of Loop 2 through a single discharge line. Normally, decay heat is transferred B-34 LM Franklin Research Center a wi.en ame r,.*w
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IQtlIIHENT IOCITION OPERATION POWER SUPPLY fM
.a Demineralized water Outside, south (true) of NA No electrical power needed 2
storage tank containment (ad jacent)
Primary water stor-Outside, about 50 ft from NA No electrical power needed age tank containment I
Frimary water PAB, ground level Control room manual and MCC B-1 (480 V) transfer pumps ALTIU nn VCT low level (both pumps) to ey Becycle PWST Outside, adjacent to PWST NA No electrical power needed i
Main condensate Turbine building, ground Main control room and Atrio pumps level on discharge header low pressure Recycle PWTPs Pall - ground level
N A*
4 8
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i
TER-C5257-311 to co.T,ponent cooling water flowirg through the shell side of the RER heat exchangers and then from the component cooling water her.t exchangers to the service water system. A secord method, prescribed in Emergency operating Procedures (EOP) 3.1-11, ' boss of Component Cooling," consists of introdacing service water directly to the shell side of the RER heat exchangere (e.g.,
under emergency conditions when comp'onent cooling water is not available). An alarm sounds in the control room if the RER flow rate drops to 2200 gpm.
RHR heat exchangers are designed for 500 psig and 400*F on the tube side and 150 psig and 200 *r on the shell side. When the RER loop is 'placed in
' service, hot reactor coolant must be introduced gradually.
Inflow is regulated with the remote-manual control bypass valve (PCV-602) while the flow indicator is observed. The temperature of the water returning from the RHR to the RCS is controlled by throttling the flow out of the RHR heet exchangers wh'ile the cocponent cooling water (or service water) flow rate is held f
constant.
RHR pumps are horizontal, centrifugal units designed for 500 psig, 400'F, and a 2200-spm flow rate at a 300-f t minimum developed head. Pump surfaces in contact with reactor coolant are austenitic stainless steel or equivalent cor:csion-rcsistant material. Mechanical seals allow nearly zero leakage of radioactive coolant to the atmosphere. Any leaks from the seals drain to the sump and are pumped to the waste disposal system.
Remotely operated McVs provide double valve isolation between the suction and discharge ends of both the RER and RCS.9ystems. Electrical interlocks
(
associated with the inboard (closest to LCS) valves prevent valve opening when RCS pressure exceeds RRR design pressure. Rey-controlled switches prevent inadvertent actuation of the outboard (closest to RHR) valves.
Redundancy Each PER pump is sized to deliver half the maximum required loop flow, and each RER heat exchanger nas half the maximum required heat removal capacity (TDSAR 5.2.3.3). Since there are two RER pumps end two RHR heat l
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1 TER-CS257-311 exchangers, a single active failure (loss of one pump or one heat exchanger) dces not completely disable the RHR system. The ut.a of two units also allows maintenance when the plant is shut down and af ter core decay heat has '
diminished.
The two methods of rcmoving heat from the RER heat exchangers also reduce the risk of loss of heat removal capacity. If necessary, a normal transfer of heat from the RER heat exchangers to component cocling water can be converted to a transfer of heat to service water introduced directly to the shell side of the RHR heat exchangers.
The RER heat Sxchanger design parameters ate given below:
Flow Tin Tout (lbm/h)
(*F)
('F)
I Shell side 1.9 x 106 95 10 5 (CCW)
Tube side l.1 x 106 140 121.8 (RCS)
Under these conditions, each RHR heat exchanger can transfer about 2,x 7
7 10 Btu /h to the CCW (or service water) system, or a total of 4 x 10 Stu/h. This power corresponds to core decay heat oi 0.65% of full power at about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> af ter the scram. The maximum design flow and heat removal rates are required when RER flow is initiated 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> af ter a scram from full power. Under these design conditions, each RER heat exchanger is capable of 7
7 removing 4.5 x 10 Btu /h (a total removal rate for two units of 9 x 10 Btu /h).
This indicates that the Haddam Neck RHR system, with an inlet temperature of 39C'F, is capable of removing a core decay heat of 1.4% of full power (corresponding to about 100 minutes af ter the scram) with both RER heat exchangers, or 0.7% of full power (corresponding to about 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> af ter the scram! with a single heat exchanger. Since it is estimatec that the use of steam venting paths for cooldown to 350'F (needed to activate the RHR system) can take up to 28 hoces before the auxiliary feedwater supply is exhausted, the RHR system heat removal capacity is considered adequate to provide uninterrupted haat remcval-following a less of either onsite or of fsite pcwer.
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TER-CS257-311 Since the RHR system has only one suction and one return line, each isolated by twe MOVs, a single failure of any valve to opan will conpletely disable the RER systen. Procedure EOP 3.1-20, " Loss of Residual Heat Removal Systen," deals with three differen*. plant situations in which the RgR system is inoperable:
(a) refueling cavity full, (b) eactor vessel head in place, and (c) reactor vessel head closure studs more than half removed. In (b),
with the reactor vessel head in place, head can be removed with the steam generators af ter the plant temperature has risen above the cold shutdown va,1ues.
In (c), with the vessel 1 Jad removed, the core can Se adequately cooled by flooding with the CVCS or other systems.
Location and Operation The staff evaluated the equipment discussed above with respect to its location and its operability during a less of offsite ac power. Table 3.2-4 I*
shows the equipment's location, the places from which it can be opetated, and itr power supply.
COMPONENT CC' LING WAWR SYSTDi Task: To remove heat from the RER heat exchangers and other essential loads.
Discussion l
The component cooling water (CCW) system is an intermediate cooling system for transfer of heat from con.ponents containing reactor coolant to the ser'. ' '.e water cooling system. During operation, component cooling water is pumped through tr.a shell side of the component coc:ing heat exchangers, where it is cooled by river water, and then flows in parallel circuits to cool the j-following componencs:
(1) reactor pumps thermal barriers and bearing oil coolers, (2) neutron shield tank cooler, (3) nonregenerative heat exchanger.
(4) seal water heat exchanger, (5) residual heat exchangers, (6) charging pump seal coolers, (7) TSR pump seal coolers, (8) boron recovery system equipment, (9) ' containment pene tration coolers, and (10) drain cooler and sample heat exchangers.
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TABI.E 3.2-4.
RESIDUAL.IIEAT REMOVAL SYSTEM i
i EQUIPMENT IDCATIOtl OPERATIOtl POWER SUPPI.Y Rllit pumps and PAB - lower level' Control room manual 400-V (MCC 5 he.it exchangers and 6) l ECS/IlllR t10Vs All 4 inside Control room manual MCC 5 (Duses 5-5 4
containment and 5-6) w Dypas.1 valve Pall - lower level Control room manual, Control air (FCV 602) locked in closed position RitR heat exchangers Pall - lower level Control room manual, Control Jir control (FCV-796) locked in open ponition b
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TER-C5257-311 The CCW loop consists of three pumps, two heat exchangers, a surge tank, cooling water lines to various components being cooled, and associated valves and instrumentation. CCW flews from the component cooling pumps, through the she%1 side of the component cooling heat exchangers, through the components being cooled, and back to the pumps. The surge line of the component cooling surge tank is connected to the suction side of the pumps. The tank, which has a capacity of 2,000 gallo?,s and a normal water level of about 1,000 gallons, is placed high enough to providt the required net positive suction head (NPSH) for propet operation of the component cooling pumps. Surge tank level is locally indicated by a level gauge and high-and-low level alarms are annunciated in the control room.
Th'e heat load on the CCW system is maximum when the RER loop is first placed in operation during plant shutdown. During normal full-power operation, one component cooling pump and one component cooling heat exchanger s*
accommodate the heat removal loads, and either of the standby pumps and a standby heat exchanger provide 100% backup.
A pressure switch on 'the component cooling pumps discharge header automatically starts the standby pumps if header pressure falls below the normal discharge level. The manual valves at the component cooling water inlet to the component cooling heat exchangers are normally open.
Component cooling water is circulated through the shell side of eacn residual heat exchanger. When the residual heat exchangers are used to cool primary coolant letdown, the ef fluent flow rate is automatically regulated to maintain the reactor coolant effluent from the exchanger at ll5'F.
This automatic central is accomplished by a temperature controller in the residual heat exchanger tube side outlet which operates a control valve in the CCW re, turn line.
Redundancy Each heat exchanger is capable of removing half the maximum heat removal load occurring between the 4th and 20th hours ef ter a normal shutdown. The use of two CCW heat exchangers and three CCW pumps assures that the heat B-40
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TER-CS257-311' removal capacity is only partially lost if one exchanger or one pump fails.
This provision also permits repair or maintenance of one exchanger and one pump while the other heat exchanger and pump are in service.
If the CCW system ruptures, 1t may be isolated from the RER heat exchangers by remote manual valves and service water substituted for CCW.
However, the remainder of the CCW system will be isolated with no provisions for cooling of components.
Location and Operation The staff evalasted the equipment discussed above with rerpect to itr location and its cperability during a loss of of fsite ac Sower. Table 3.2-5 gives the equipment's locati3n, the points from which it may be operated, and s'
its power supply.
SERVICE WATER SYSTD1 Task: To supply cooling water to the CCW system and RHR heat exchangers.
I Discussion l
The service water system (SWS) censists of four pumps which supply water from the Connecticut River to a dual header system in which two parallel full-size her.dcrr supply bc*.h the primary an3 secondary plants.
In the turbine building, each headar divides into a primary supply and a secondary supply header. Power operated valves at the beginning of each secondary plant header l
l automatically rhut to secure secondary service water if of fsite power is lost, and the SWS is reduced to two pumps supplied by the diesel generators [4].
1-Similar provision is made for shutting off nensssential supplies to primary I
plant equipment. Also, remote manual valves can be used to substitute SWS l
l flow for CCW syst0m flow to e.Se PCR heat exchangers. Each SWS branch -
connecticn to a system heat load in both the primary and secer.dary systems is connected to both headers, and valves permit shutof f from either or both headers. During normal operation, ooth header systems operate in parallel.
t j
The SWS pumps and the valves which switch the RER heat exchanger cooling i
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COMPONENT COOLING WATER SYSTEM EQUIPMENT IDCATION OPEttATION POWER SUPPLY 4
CCW pumpri (3) pan Control room manual.
P-13A 480V nus 4 Auto on low discharge header.
l'13B 480V Bus 6 Pressure.
P-13C 480V Dus 7 m
1
" CCW heat exchangers PAB TCV-ll2 (NRHX)
PAH Local manu.'.1 control.
Air-operated (falls open)
Automatic temperature control.
ICV-608 (MCP scal) pan 7
Air-operated
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TER-C5257-311 supply from either the SWS or the CCW system are controlled froni the control room. Power for the pumps is supplied from the 480-V emergency buses which can be powered from ensite or of fsite sources.
Redundancv During the shutdown and cooldown following the loss of ac offsite pcwer, two service water pumps are powered from the emergency power supplies. Upon the loss of all normal ac power and af ter the emergency power supply is established, one service water pump will be started automatically on each diesel genernce.
If the first service water pump does not start,' the power supply is automatically transferred to.the second pump on the diesel generator bus.
I Operation and Instrumentation The staff evaluated 'the location of the various equipment discussed in this section and the relevant instrumentation available to the control room cperaters. Table 3.2-6 lists the equipment, its location, the places from which it can be operated, and the equipment's pcwer supply.
CEMICAL AND VOLT 2E CON *RCL SYSTm l
Task: To provide RCS makeup (needed because of coolant contraction during cooldown), to borate the RCS to the necessary shutdown margin, and to j
provide auxiliary pressuri:er spray.
Discussion The chemical and volume control system (CVCS) consists of two centrifugal charging pumps, the regenerative heat exchangers, pressure reducing valves and orifices, either or both of the two RHR heat exchangers, the volume control tank (VCT), refueling water storage tank (RWST), and associated' piping, valves, fittings, and instruments.
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[S-2 kQ TABI.E 3.2-6.
SERVICE WATEft SYSTFM Ia 3
TQUIPMEMP IOCATION OPERATION POWER SUPPIY i
Service water pumps Screen house Manual start from the P-37-1A 480-V Dus 4 control room; automatic P-37-1B 480-V nus 5 start upon loss of ac power P-37-1C 480-V Dus 6 i
and subsequent emergency P-37-lD 480-V Bus 7
?
diesel generator startup a
SWS header 2 - Turbine b1dg. (Mov-1,-2)
Automatic upon loss of ac MCC-5 isolation v.'Ives 4 - PAB (MOV-3, -4 ;
power and manual operation I
ADV-8,-9) f rom control room Buses 5-5 and 5-6 i
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During normal operation, reactor coolant is withdrawn from the cold leg of Loop 1.
The coolant then passes through the shell side of the regenerative heat exchangers (RHXs), the three letdown orifices in parallel, and then through the tube side of the non-r,egenerative heat exchanger, af ter which the coolant temperature and pressure have been reduced to abcut ll5'F and 200 psig, respectively. Af ter further pressure reduction to 150 psig, the coolant flows through various demineralizers and into the VCT.
The charging pumps'ta'ke '5tcticn from the VCT, then pump the coolant through charging and pressurizer level control valve's, through the tube side of the RHX, and then in'to Loop 1.
Centrifugal types with a capacity of 360 gpm at 2300 psig, the charging pumps can take suction from the folicwing sources:
1.
RWST (gravity flow) 2.
VCT (gravity flow) 3.
chemical blender (BATP) 4.
boric acid tank (BATP).
During periods of decreasing power level, the charging flow is increased by the pressurizer level centrol to make up for the contraction of reactor coolant water which consequently decreases the temperature of the letdewn stream leaving the regenerative heat exchar.ger. There is sufficient space in the VCT between the makeup set point and radioactive waste disposal system dump set point to accommodate coolant expansien and centraction resulting from changing reactor power level, but makeup is available to compensate for any leakage that may have occurred during power operation.
Boron concentrations are controlled by the chemical blender, boric acid
. transfer pump (BATP), and reactor makeup control system which act together to l
inject a predetermined volume of mixed boric acid and pure water into the VCT.
There are three modes of operation: automatic makeup mode, dilute mode, and borate mode.
Beration of the RCS with the blender uses the 12,000-gallen tank of 8%
boric acid. The refueling water storage tank can supply the charging pumps directly (230,000 gallons of water with a boron cencentration of at least 2470 l
l 1
E-4 5
. 2 Franklin Research Center m.aemev e.u.
TER-C5257-311 ppm). The chemical blender and makeup control system are normally utad for RCS boration during shutdown and cooldown.
Auxiliary pressurizer spray is also provided by the CVCS. Upon less of offsite power and reactor trip, main coolant pumps (MCPS) will stop due,to insufficient electrical power, which, in turn, will remove the source of pressurizer spray (MCP differential
- pressure). Auxiliary pressurizer spray from the charging system can be initiated to assirt in plant pressure control during hot shutdown and plant depressurization during cooldown.
,Redundanev The amount of RCS makeup during cooldown from 544'T to 200'F and filling of the pressurizer was calculated by the staf f to be about 18,900 gallons, beycnd the 12,000-gallon capacity of the' boric acid tank alone. Other sources of primary grade water are available (i. e., VCT, RWST).
I Cooldown calculations were performed to verify that the pressurizer level can be controlled during the most rapid cooldown. This cooldown rate was initially (i.e., at T
= 544*F) slightly greater than 100'F/h, the RCS Technice'. Specifications maximum (during a 1-hour interval). -A cooldown rate of 100*F/h causes an RCS liquid contraction rate of about 133 gpm. Since the capacity of each charging pump is about 360 gpm, the pressurizer level can be raised by only one charging pump during a cooldown while venting steam from the steam generators, and the other charging pump provides a redundant RCS makeup capability.
The 30-gpm metering pump can also be used as an alternate means for injecting berated water into the RCS.
Since the metering pump does not have sufficient capacity to restore the necessary coolant make-up volume during cooldown, it is not considered part of the equipment associated with the minimum systems list. The metering pump has been addressed because it does provide a path for borated water injection to the RCS.
The Haddam Neck Technical Specifications state that 10,000 gallons of an 8% solution of boric acid are required to meet cold shutdown requirements.
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TER-CS257-311 The 12,et0-ga11<m boric acid tank with a minimun. ?% solution is sufficient to provide the required shutdown margin. Similar plants have used the RWST in place of the boric acid tank and blender in the minimum systems list.' The RWST is a source of primary g'rade water with sufficient volume (230,000 gallons) as required in the Technical Specifiesticns to make up for plant e
centraction during cooldewn (19,000 gallons) and sufficient boric acid concentration to achieve the required shutdown margin. Therefore, the RWST
'h a: been included and the boric acid tank omitted from the list of equipment associated with the minimum systems list.
Loss of control air will isolate CVCS letdown flow from the RCS, Letdown flow will require n.anual initiation, manual flow rate control, and manual temperature control (via manual adjustment of CCW flow to the ncn-regenerative l heat exchanger) in order to borate the RCS before cooldown. In addition, the beric acid blender isolation valve will feil shut, requiring manual operation to provide a backup supply from the boric acid tank.
Location and operation f
The staff evaluated tne equipment discussed above with respect to its l
location and its opacability during a loss of of fsite ac power. Table ? 2-7 gives the equipment's location, the points from which it may be operated, and its power supply.
PRESSURE CON"'RCL AND RELIEF SYSTD4 Task: To maintain reactor coolant pressure af ter shutdcwn while in a hot standby or a natural circulation cooldown mode, and to depressurize the main
" coolant system to permit initiation of the RER system.
Discussion The pressure centrol and relief system maintains the required reactor coolant pressure during. normal s,teady-state operations, accommodates changes l
in system volume, and limits changes in system pressure caused by power i
transients. Following reactor shutdown, the pressure control und relief l
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TABI.E 3.2-7.
CIllMICAL AND VOLINE CONTROL SYSTIM ME
- t. a fy FQUI PMEhT IOCATION OPERATION POWER SUPPLY 0
{p}
Charging pumps (2)
Separate cubicles on the Operable from the control CPA:
4-kV Bus 9 In 21'6" level of the PAB room and AlfrO star t on low CPB:
4-kV Bus 8
!j discharge llDR pressure Volume control Atove charging pumps on ' the Makeup to the VCT via the No electrical power needed i
tank 35'6" Ic el of the PAB makeup control system controlled from the con-trol room RWST SE corner of the PAB, just
' Level instrumentation and m
north of containment makeup control f rom the 1
control :oom. Manually I
a ligned in the PAD Doric acid tank 35'6" level of the PAD Level instrumentation is in Tank agitator: MCC 8-1 Bus U
(BAT) the control room; local 8-5; IIcat tracc: MCC 8-1 makeup control is manual in Guses 8-5 and a-G the PAB i
i Boric acid transfer Beneath the BAT, 21'6" Operable from control room BATP-1A: MCC 8-1 Bus 8-5 i
pumps (BATPs) level of the PAB and AUTO start on VCT low BATP-1B MCC 8-1 Bus 8-6 I
level D3 11 "xs A
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TER-C5257-311 system responds to maintain reactor coolant pressure during hot shutdown and cooldewn. Coolant pressure control is necessary to maintain an adequate subcooling margin and to prevent disruption of either forced or natural circulation. The pressure centrol system also provides means for t
depressurizing the RCS to allow operation of the RER system.
The pressure control and relief sys:em consists of a pres'surizer vessel containing a two-phase mixture of steam and water, immersion heaters, spray systems, cafety and relief valves, and associated piping, valves, and inr trumenta tion. The 3-inch spray line is connected to the top of the pressurizer a.nd receives flow frcm the cold leg of Loop 3 or Loop 4 or from the auxiliary spray system (CVCS). The 10-inch surge line connects the bottom of the pressurizer to the Loop 4 hot leg. Pressurizer heaters are used to l centrol pressure changes af ter shutdown to make up for ambient or cooldown haat losses. The heaters are powered frem non-safety-related buses as follows:
l.
centrol heater group C - Bus 5
~
2.
backup heaters group A - Bus 4, group B - Bus 5, group D - Eus 6, and group E - Lus 7.
If a, power less disables the pressurizer Fe*'-ta, pressure may be controlled by charging to the RCS directly from the CVCS cr by filling the pressurizer and controlling plant pressure with CVCS charging pumps while in a solid water condition.
The RCS is normally depressurized by means.of the 3-inch pressurizer spray line f cm either Leop 3 ce Loop 4.
Normal driving head fcc pressurizer spray flew is derived from head losses associated with the reactor vessel anti steam generator. However, a loss of either ensite or of fsite power would prevent the main coolant pumps from sustaining coolant flew and, consequently, the driving head necessary for main pressurizer spray. As an alternative, auxiliary spray from the CVCS via McV-299 can be used to supply pressurizer spray flow and centrol pressurizer pressure.
The pressurizer safety and relief valves can also be used to reduce coolant pressure by venting steam from the top of the pressurizer, drawing A. -
B-49 x.; Franklin Research Center w anor m,w,
i TER-C5257-311 colder water into the pressurizer via the surge line, and reducing pressurizer T'e three safety valves are designed to prevent temperature and pressure.
a system pressure from exceeding the design pressure of the pressurizer and the reactor coolant system by more than 10% and the power-operated relief valves (PORVs) are designed to operate at a lower pressure (2270 psig) to limit the operating frequency of the safety valves. The PORVs are solenoid-controlled, air-operated valves which may be operated from the main control board. Each pORV may be isolated downstrer, c' The valve by a remote motor-operated block valve.
The pressurizer relief t C eceives the discharges from the PORV and safety valves and cools the.a..Aowing water while prever. ting contamination of the containment. It is designed to permit condensation and cooling of a pressurizer steam discharge while the tank temperature rises from 120*F to 2 0,0
- F.
If its temperaturt exceeds 120*F during normal plant operation, the Ipressurizer relief tank may be cooled with spray from the primary water storage tank while the warm mixture is drained. Spray and draining operations are controlled from the main control board.
Redundanev Lass of offsite power causes various consequences:
a.
Main coolant pumps and pressurizer heaters, powered from non-safety-related buses, are deenergized.
b.
The step reduction in load leaves residual heat in the core and reactor coolant system, which raises reactor temperature and pressure.
c.
The cessation of main coolant pump flow removes the source of normal pressurizer spray.
d.
Deenergizing of the pressurizer heaters disables the norcal means of pressure control.
Pressure can be reduced by various means after loss of offsite power and a step load reduction. Coolant flow front the CVCS can be diverted through MOV-298 to the pressurizer spray inlet line to supply the auxiliary spray system which has a capacity of 200 to 250 gpm.
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TER-C5257-311 As a backup, the safety vals as and PORVs are designed to accommodate pressure surges beyond the normal pressure limits of the pressurizer and the spray system. Since auxiliary spray will not be available until emergency power is supplied to the charging pumps, the safety valves and PORVs provide the interim means of pressure reduction. Two PORVs (568, 570) provide redundant depressurization paths. Both valves are powered from MCC-5 and operate to limit RCS pressure transients to less than 2270 psig. Both PORVs are
' air-operated, solenoid-controllec valves, requiring both compressed air and electrical power to operate, and fail shut on loss of of fsite power.
Electrical power is automatically restored (in approximately 10 seconds) when the emergency diesel generators are started; however, the control air compressor must be nanually reenergized by the operator af ter power is restored in order to supply compressed. air to operate the valve.
?
Downstream of each PORv is a motor-operated block valve (567, 569) which is normally closed and fails shut on loss of power. These valves are powered from the same motor rentrol center (MCC-7) a s th e PO RVs.
If the pressure transient exceeds the PORVs' capacity, three safety relief valves auto-metically open to preven
- the pressure from rising more than 10% above the systea's 2485-psig design limit. The maximum capacity of the safety relief valves permits them to _elieve the pressure transient caused by a complete loss of J oad without r'eactor trip.
l As a result of loss of offsite power, all five heater groups are tripped and locked out, and must be manually' reenergi:ed by the operator when the j
emergency power is available from the diesel generators. The Licensee has stated [9] that a Westinghouse study was performed which determined that 118 kW were required to compensate fer ambient helt losses and maintain a suitable margin to saturation conditions to keep the RCS adequately subecoled and allow natural circulation to occur. The vendor study also recommended that a bank of backup heaters be available to each emergency power train within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after a less of offsite power (backup heater groups A and I, with 305 kW each, are utilized), although the indicated loss of subcooling (and, as a result, nataral circulation) would not occur 'for 5-6 hours if no heaters are t
l l
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t MU Franklin Resesrch Center B-51 4 Cansson of The F aeme testMe e
e TER-C5257-311 available. EOP 3.1-9, " Total Ioss of AC," directs the operator to "raintain coolant pressure above saturation wita the pressurizer" af ter emergency power becomes available to the 480-V buses.
If pressurizer heaters are not operating, an alternate mear.s of pressure control is the discharge of pressurizer steam volume while filling the cressurizer 'with the CVCS charging pumps. This cools the pressurizer, collapses the steam bubble, and fills the pressurizer with water. System pressure is then maintained in a solid water condition by intermittent operation of the positivc displacernent charging pumps. Two charging pumps are available, powered from separate emergency buses, and can be ' lined up independently to supply coolant t<>
(and therefore maintain pressure in) the reactor coolant system. The Licensee states that there is no explicit '
~
pro. edure available for operation with a, solid plant other than those steps included in NOP 2.3-4, " Plant Shutdown-Hot Standby to Cold Shutdown."
J Location and Operation The staff evaluated t'he equipment discussed above with respect to its location and its operability during a loss of offsite ac powez;. Table 3.2-8 gives the equipment's location, the places from which it may be operated, and the equipment 's power supply, CCb"lROL AIR SYSTD!
Task: Provide compressed air for instrumentation and the control of I
l air-operated valves in other safe shutdown syr: ems.
l Discussion:
"vo 166-sefm control air compressors deliver 100-psig air to the instrument and control air system. The compressors are vertical, I
single-stage, direct-acting, water-cooled, non-lubricated carbon ring types, each belt-driven at 514 rpm by a 40-hp motor and equipped with an after cooler, an air receiver, and an air intake filter. Each compressor has the capacity to meet the full plant requirements for instrument and control air, and both receive their power from a different station service bus section.
O B-52 2ap Franklin Research Center A Devon of The Freamhn ansecute
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5h T3nLE 3.2-8.
PRESSilRE CONTROI AND RELIEF SYSTEM a
EOllTPMENT IDCATION
' OPERATION PO* DER SilPPLY Pressurizer heaters In pressurizer Control room
- MCC Control group C 480V Dus 5 (automatic and manual)
- Backup group A 480V Bu.s 4 group B 480V Bus 5 group D 480V Dus 6 group E 480V Dus 7 fn
- Power-operated In co.itainment Control room EC-5 Control air solenoids relief valves near pressurizer Auxiliary spray In containment Control room MCC-5 nus 5-5 control valve (elev. S'6")
Block valves In containment control room MCC-5 connectable to either 480V Bus i
(ab<we 4 8 6")
5 or Dus 6 I
- Pressurizer heaters are locked out from normal power supplies on loss of offsite a
power. On loss of of fsite power, buses 4 and 5 arc powered from diesel generator E
bus 2A, buses 6 and 7 are powered from diesel generator bus 2B.
A us W
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k i
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TER-C5257-311 Each compressor has dual control and is provided with a local control switch with "Off-Eand-Auto" positions.
In the " Hand" position, the compressor runs continuously wi.th the compressor valves loading and unloading to maintain receiver pressure within the established limits.
Ir the " Auto" position, the compressor motor is started and stopped automatically to maintain receiver pressure. In the "Of f" position, th9 compressor motor is shut down. While one compressor operates, the secor.d, a stand-by compressor with its control switch in either the " Hand" or " Auto" porition will automatically supply air, or will start and supply air if the system air pressure f alls below the control range of the operating compressor.
The control air supply in the pressure range of 90 to 100 psig is divided between ti.e two compressors. A transfer switch allows the higher load to be periodically transferred from one compressor to the other,. thereby reversing' operating and stand-by compressors.
I*
Control air is supplied to the plant through a double piping system originating at the two control air receivers. Air from each receiver passes through a prefilter, a depydrator, and an af te.r-filter in series, each with a rated capacity of 168 cfm.
Dehydrators are of the silica gel, automatically regenerated type. Each is in turn divided into half-size sections. This arrangement provider 100% to 150% of full control and instrueent air supply while one dehydrator section is being regenerated.
l Double valved connections are provided at several points between the two l
parallel control air piping system?.
In general, local panel boards and air supply headers to groups of instruments are fed through local filters and througn pressure regulating valves from each of the two mein parallel air supply systems.
Redundancy Following a less of offsite ac power, the control air compressors would nct be available since they are powered from separate non-emergency 480-V ac buses. The operator can re-energize the non-emergency buses when power-is
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_nklin Resea_rch Center
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TER-CS257-311 available to the emergency buses. " Total Loss of AC," ECP 3.1-9, however, does not direct the operator to perform this action. ECP 3.1-34, " Complete Loss of Control Air Supply," describes the immediate and subsequent operator actions required to place the plant in a hot shutdown condition.
Loss of of fsite ac power and subsequent loss of centrol air will result
~
in the 'following significant actions in related safe shutdosn systems:
o FCV-202, 203 and 204 (CVCs letdown control valves) will close an'd isolate letdown flow from the RCS.
Letdown ficw is required to borate Las RCS.
.o FCV-llo and 308 (CVCS charging control valves) will fully open to provide a charging path to the RCS.
o FCV-ll2B and ll2C (CVCS boric acid blender isolation valves) will close, isolating flow to and from the boric acid blender.
o FCV-6 02 (RHR heat exchanger bypass valve) will close and FCV-796 (RHR heat exchanger outlet valve) will,cpe n.
Manipulation of these two valves is required to initiate RHR flow for long-term cooling.
o TCV-ll2 (CCW temperature centrol valve to the non-regenerative heat exchanger) will open.
o PR ACV-568 and 570 (pressurizer POKVs) will close, isolating one of the means of depressurizing the reactor coolant system to a level that will allow RER initiation.
HICV-13 01- (1-4') (f eedwater regulating valie bypass valves) and o
FRV-13 01- (1-4 ) (feedwater regulating valves) will close, isolating feedwater flow to the steam generators. Manual operation of the bypass valves is required following loss of offsite ac power to reinitiate feedwater flew to the steam generators.
In addition, the bypass valves will require repositioning to maintain a suf ficiently high steem generator water level.
o HICV-1201 (main steam atmospheric dump valve) will close and cannot be manually repositioned.
o PICV-1206A and 1206B (main steam isolation valves for auxiliary feodwater turbines) will closs, isolating the turbines and requiring manual cperation to start the turbines.
o Pneumatic level-indicators, including pressurizer level, will give erratic or erroneous indications.
B-55
.a Franklin Research Center
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TER-C5257-311 An alternate source of pressurized air is the service air system, which can supply the control air system through a tie provided with a regulating valve set at 70 psig. The service air compressor is also powered from non-amergency buses and would not be availatle af ter a loss of offsite ac power.
The control aix system supplies pressurized air for necessary valve control functions within safe shutdown systems and for pneumatic signals in essential instrumentation. Loss of the control air system will not prevent the plant from achieving a safe shutdown condition but ir detrimental because it necessitates manual operation of numerous valves in other safe shutdown systems. These actions are additional to any limited operator actions which may accrue from e single active failure. On the basis of the above discussion, the staf f concludes that the control air system does not satisfy the functional requirements of BTP RSB 5-1 in that a reliable source of bontrol air is not availalle and significant operator action outside the control room is required to effect a safe shutdown. The staff will evalcate the significance of this in the SEP integrated assessment of Haddam Neck.
Location and Ooeration The staff evaluated the equipment discussed above with respect to its location and its operability during a loss of cffsite ac power. Table 3.2-9 gives the equipment's location, the points from which it may be operated, and its power supply.
EVIRGENCY POWER SYSTIMS AND INSTRW2NTATION Task:
To supply a reliable source of ac and dc ;ower to necessary equipment and to provide sufficient instrumentation for control of equipment functions.
Discussion The staff evaluation of emergency power systems and instrumentation,
~
their reliability, operability, and associated electrical distribution will be conducted later under several S'EP topics.
B-56
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i TABLE 3.2-9.
CONTHOL AIR SYSTEM 10tlI PMEfff II) CATION CONTROL POINTS ELECTRIC POWER SUPPLY Air compressors (2)
Turbine building Local (in Auto, automatic C-3-1 A MCC 5-1 Bus 5-5
.j y
(ground grade at mrth start of S'IBY compressor C-3-IV HCC 5-1 Bos 5-6 end of auxiliary bay) on low system pressure)
Air receivers (2)
Same as alove e
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TER-C5257-311 3.3 SAFE SHUTDOWN INSTRUMENTATION Table 3.3 lists the instruments required to conduct a safe shutdown. The list includes those instruments that indicate information such as RCS pressures and temperatures and pressurizer and steam generator levels that enable the control room operator to judge whether all safe shutdown systems are operating properly.
Improper trends in these indications would alert the operator to investigate possible causes. Other instruments listed in the table provide the operator with (1) a direct check on safe shutdown system performance, and (2) an indication of actual or impending degradation of system performance. This list of instruments satisfies the requirement of BTP RSB 5-1 for safe shutdown. The DBE evaluations, which in many cases are not based on the same asrumptions as this review', may determine that additional instrumentation is required to achieve and maintain a safe shutdown following
^
a DSE.
Designs of the instrumentation and controls used fee safe shutdown t'ill be evaluated later in the electrical portion of the resolution of SFp i.
Topic VII-3.
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.M'.' Franklin Research Center
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TABLE 3.3 SAFE SilUTDOWN INSTRUMENTS u
- a. 3 Component / System Instrument Indication Instrument Location, Control Room Indication fE ly.
Reactor Coolant System Pressurizer Level In containment yes In Pressurixer Pressure Irl containment yes l
4 ItCS T(.mperature In containment yes Steam Generator Steam Generator Level In containment yes Auxiliary Feed System Demineralized W ter At. tank yen Storage Tank Level Primary Water Storage At tank
?
m Tank Level Auxiliary Feed Flow Hate Chemical and Volume' VCT Level At tank yes Control System BAT Level At tank yes RWST Level At tank yes In PAB
?
CVCS Pump Flow Hate Itesidual lleat Removal RilR System Flow Itate In PAD Alum on' Low flow System FCV-602 (Dypass Valve) yes Component Cooling Water CCW Syst'em Flow Itate In PAB yes System CCW Surge Tank Level In PAB
?
CCW Temperature
?
yes O
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- I:rn l "R Component / System Instrument Indication Instrument Location Control Room Indication 3
Service Water System Service Water Flow Rate In screen house Pump amperaqe only Diesel Generator Generator output 7
yes (voltage, current, frequency, power)
Emergency AC Power 4160-V Dunes IC and 2C 7
yes Voltage Indication and o
Iow Voltage Alarm 400-V Switchgear yes Voltage Indication and Inw Voltage Alarm 120-v Buses yes Volt Indicating Lights, Alarms, voltage, and Frequency Emergency DC Power 125-V DC yes o
Undervoltage Alarm y
Dus Ground Alarms U
v.
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TER-C5257-311' 4.
SPECIFIC RESICUAL HEAT RDIOVAL AND OTHER REQUIREMENTS OF BRANCH TECHNICAL POSITION 5-1 Branch Technical Position 5-1 contains detailed functional requirements for specific systems used during safe shutdown. Each specific requirement is presented belew with a description of the applicable Haddam Neck system or 4
area of operation.
~.
4.1 RER SYSTDi ISCLATICN REQUIREMEN.S Recuirement:
"B.1 The following shall be provided in the suction side of the RER system to isolate it from the RCS.
(a)
Isolation shall be provided by at least two power-cperated valves in series. The valve positions shall be indicated in the control room.
(b).The valves shall have independent diverse interlocks to prevent the valves from being opened unless the RCS pressure is below the RER system design pressure. Failure of a power supply shall not cause any valve to change position.
~
(c) The valves shall have independent diverse interlocks to protect against one or both valves being cpen during an RCS incret..se abcVe the design pressure of the RRR system."
Evaluation:
(a)
Haddam Neck PSR station line has two pcwer-operated valves in series.
Pesitions of both valves are indicated in the control rocm.
J (b) The upstream (i.e., closest to the RCS) valve, MOV-780, has an interlock which prevents its opening unless RCS pressure (as sensed by the four RCS pressure channels) is less than 400 psig. The other (downstream) valve, Mov-781, is under administrative contrels only. Both motor-operated valves fail "as is" on loss of power.
(c)
Neither of the two suction valves has an interlock to protectively close an open valve when the RCS pressure rises above the des'ign pressure of DA B-61 f.N Franklin Research Center a Oms.on W N Franen instAge
.-nn.
4 TER-C5257-311 the RER system. However, the overpressure protection system (OPS) includes a 400-psig alarm to warn the operator that RCS pressure is increasing, enabling the operator to terminate the pressure increase or isolate the RHR system.
Recuirement:
"B2.
One of the following shal1 be provided on the discharge side of the RRR system to isolate it from the RCS:
(a) The valves, position indicators, and interlocks described in Item 3.1 (a)-(c).
(b) One or more check valves in series with a normaily closed power-eperated valve with its position indicated in the control room.
If the R".R system discharge line is used for an ECCS function, the power-operated valve should be opened upon receipt of a safety injection signal once the reactor coolant pressure has decreased below the ECCS design pressure.
(c) Three check valves in series, or (d) Two check valves in series, provided that there are design provisions to permit periodic testing of the check valves for leak tights.ess and the testing is performed at least annually."
Evaluation:
The Haddam Neck RER system has two discharge paths which require isclation from RCS pressure:
the normal RER return to the RCS (via two power-operated isolation valves), and the discharge to the ECCS core deluge supply lines (via a single check valve in series with a power-operated isolation valve).
The normal RHR return to the RCS must meet the criteria of Item 2(a).
Features related to compliance are as follows:
l.
This discharge line has two power-operated isolation valves, MOV-804 and MOV-805, and positions of both valves are indicated in the control room.
2.
An interlock prevents the upstream valve, MOV-804, from opening until RCS pressure (as sensed by four RCS pressure channels) is less than 400 psig. The downstream valve, MOV-805, is under administrative controls only. Both valves fail "as is" on loss of power.
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M TER-CS257-311 3.
Neither of the two discharge valves has a protective interlock to protectively close an open valve when the RCS pressure rises abc 2 the design pressure of the RER system. However, the OPS includes a 400-psig alcrm to warn the operator that RCS pressure is increasing, enabling the operator to terminate the pressure increase or isolate the RHR system.
The two RHR core deluge discharge paths, each with a sinele check valve (872A, 8723) and a power-operated isolation valve (MOV-8711, MOV-871B), must meet the criteria of Item 2fb).
Features related to compliance are noted as
'folicws:
The core deluge line from the RHR is part of the flow path from the low pressure safety injection (LPSI) pumps supplying long-term low pressure ECCS recircul2 tion flow. Following a postulated icss-of-coolant accident (LOCA),
after 100,000 gallons of water have been injected by the safety injection (SI) f system, the RHR system is started and aligned to take suction on the containment sump to re;ycle and cool the spilled reactor coolant. The core deluge paths branch from the RHR discharge line, which splits into two parallel lines, each containing a check valve and a power-operated isolation valve. The positions of the power-operated valves (MOV-871X and MOV-871B) are displayed in the control room. The power-cperated valves open i= mediately upon receipt of the SI signal. The check valves are periodically leak-tested.
Conclusien:
On the basis of a comparison of the above description of the RHR system with BTP 5-1 provisions 2(a) and 2(b), the following deviations are noted:
1.
The RER-to-RCS suction ar.d discharge isolation valves do not have independent diverse interlocks to prevent the i
(
valves from being opened unless RCS pressure is below the RHR system design pressure.
2.
The RRR-to-RCS suction and discharge isolation valves do not have independent diverse protection to close an open valve if RCS pr4 sure rises above the RHR system design pressure.
3.
The RHR-to-core deluge isolttion valve epens immediately upon receipt of an SI signal'.
I -63
..'.dd Franklin kesearch C< enter 4 N oa t N r.,wan in. wie
TER-C5257-311 l
The staff has concluded the following:
1.
The Licensee *must correct the lack of independent diverse interlocks for the isolation valves to prevent the valves from being opened unless the RCS pressure is below the RER system design pressure.
2.
Tne lack of valve closure if RCS pressure rises above RRR system design pressure is acceptable because it is compensated by administrative and procedural controls of i
these valves and the warning alarm provided by CPS.
~
i 3.
The acceptability of the opening of power-operated valves (with ECCS functions) 'upon receipt of an SI signal before 3
RCS pressure f alls below ECCS design pressure will be evaluated in the integrated assessment.
I i
4.2 PRESSURE RELIEF
,QUIREMENTS
.he RHR system shall satisfy the following pressure relief requirements.
Recuirement:
"C.1 To protect the RHR system against accidental overpres-surization when it is in operation (not isolated from the RCS), pressure relief in the RER system shall be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code. The most limiting pressure transient during the plant operating condition when the RHR system is not isolated from the RCS shall be censidered when selecting the pressure relieving capacity of the RER system. For example, during shutdown cooling in a PRR with 'no steam bubble in the pressurizer, inadvertent operation of an additional charging pump or inadvertent opening of an ECCS accumulator valve should be considered in selection of design bases."
Evaluation:
- The RHR relief valve has a setpoint of 500 psig and a relief capacity of 960 gpm [10). This r211ef valve was not sized to accommodate the highest j
pressure transients postulated to occur during RHR cooling of the RCS.
However, the Licensee analyzed these potentially most severe pressure
-transients during the NRC generic review of RCS overpressurization events
[11]. To prevent or mitigate these transients, the Licensee has mad'e several orocedural and herdware modifications. The hardware modifications are the D'
B-6 4 0.15 Franklin Pesearch Center A Des.on o The Franen easienne r
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TER-C5257-311' CPS, which relieves RCS pressure, with one exception, by means of spring safety valves connected to the pressurirer and is designed on the assumption of a water-tolid RCS and the most limiting single active failure. The exception is a pressure increase dJe to a mass input by the high pressure safety injection pump, for which the Licensee has implemented administrative controls. When the RER system is cooling the RCS', the CPS also provides overpressure protection fcr the RER system.
The effects of the worst cases of mass and heat inputs were evaluated to
' establish the capability of the CPS and RER relief to prevent RER overpres-r;rization. The mass input case is presented in Reference 9 and the heat input analysis follows.
For the heat input case, the data supplied and referenced (12] by the lL.icensee ware extrapolated to include a 50*F temperature difference between the steam generator and the RCS at an RCS temperature of 300*F (the referenced data applied.only to heat input transients associated with RCS temperatures from 180*F to 250'F),
The 300*F temperature was chosen because this is the temperature at which the Licensee initiates RER cooling.
The staff determined that pressure transients due to the worst-case heat input would not exceed 110% of AER design pressure if both the CPS and RER relief mechanisms functioned. They also determined that RER overpressure woJ1d not occur at an RCS temperature of 200*F even if one of the three valves failed.
The staf f then considered the potential for a heat input transient at Haddam Neck when the RCS temperature is between 250*F and 300 *F.
The occurrence of such a transient would require a rapid transfer of heat frc: the steam generators to a cooler RCS in a water-solid condition by forced convection caused by a reactor coolant pump start. In its review of l
overpressurization transients, the staff considered a diffei.nce of over 50*F l
l between the steam generator and RCS temperatures to be unlinely. The 1
administrative measures proposed by the Licensee to reduce the probability of heat impact transients were to (1) minimize p.
..t operation in a water-solid condition, (2) limit the difference between the steam generator and RCS B-65 JA') FrankJin Research Center A o-on om rma.n m.m.
L
TER-CS257-311 temperatures to no greater than 50'F, and (3) prohibit reactor coolant pump start when the ster.m generator temperature is above the ROS temperature.
Although measures 1 and 2 would not necessarily preclude a heat inp"t transient, u.aasure'3 would. Also, the staff assessed the possibility of a heat input event during plant cooldown, which is the time when the ~ steam generntor tedperature may exceed the*ROS temperature while the RCS temperature is above 250'F.
The Licensee activates the PRR system af ter cocling down to 300*F with the steam generators. Continuing the cooldown with the RER system and with reactor coolunt pumps secured would result in the development of the 50*F differen -e at an RCS temperature of 250*F.
As noted before, a heat input event a ; this temperature would not overpressurize the FER system even with an asenmed single failure. The staff's evaluation of the OPS and associated Technical Specifications was issued in Reference 13.
l Cenclusion:
The CPS and PRR reliefs meet the pressure relief requirements of BTP 5-1 in that:
1.
the CPS and FRR reliefs provide sufficient F3R overpr' essure protection during the postulated worst case mass input 2.
the CPS and RER reliefs provide sufficient EER overpressure protection at RCS temperatures of 250*F c less. Above 250'F, the Licensee's procedure acceptably minimizes the likelihood of a heat input overpressure transient.
By Technical Specification, the CPS (or equivalent vent area) must be operable if RCS temperature is less than 340*F.
At present, the Licensee has proposed a procedure for placing the OP! in operation before initiating RER system cooling. The staff will conside, in the SEP integrated assessment of Haddam Neck, making this OPS /RHR seque.ce part of the facility Technical Specifications.
Recuir e med t:
"C.2 Fluid discharged through the RHR system pressure relief valves must be collected and contained such that a stuck open relief valve "ill not M
B-6 6 DJ Franklin Research Center
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__._____A__
e TER-CS 257-311 (a)
Result in flooding of any safety-related equipment.
(b)
Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA.
(c)
Result in a non-isolatable situation in which the water provided to the RCS to maintain the core in a safe condition is discharged outside of the containment."
Evaruation:.
(a)
Fluid discharged through the RER relief valve is directed to the refueling water storage tank (RWST) and cannot flood any safety-related equipment.
(b) The RHR relief is connected to the discharge of the RHR pumps and, I if the valve ?hould stick open, core deluge flow and post-LOCA recirculation flow would be affected. The RER relief valve flow rate is 960 gpm at a relieving pressure of 500 psig.
In a post-LOCA scenario, the pressure experienced at the RER relief will be either LPSI pump discharge pressure, if core deluge ficw is being delivered, or RER pump discharge pressure, if recirculation is in pregress. As noted in the Technical Specifications [14),
only cne LPSI pump, with a flow rate of 5500 gym at a discharge pressare of 254 psig, and only one RER pump, with a flew rate of 2200 gpm at a df.scl.arge pressure of 130 psig, are required to be operable to supply 100% of core deluge and recirculation requirements.
The Haddam Neck ECCS is provided with two LPSI and two RER pumps to meet the redundancy requirements posed by the single f ailure criter *cn of the ECCS Interim Acceptance Criteria. Therefore, the leakage through the RER delief (less than 960 gpm) would reduce ECCS flow less than would the loss of an LPSI pump (5500 gpm) or an RHR pump (2200 gpm).
(These losses have been postulated as single failures in the ICCS analysis.)
The Technical Specifications requirement for ECCS pump operability at Haddam Neck requires only one train of ECCS to be operable whr.never the reactor is critical (ene train of ECCS includes one RER pump, one LPSI pump, 4
B-67 t
... 9 Frank'in Research Center acu aw 3.r= = am m.
[
TER-C5257-311 one EPSI pump, and one charging pump). However, no Technical Specifications requirement governs the allowed outage time of the other ECC3 train. Such a requirement should exist to maintain the ECCS redundancy assumed in the DCCS analysis. This Technical Specifications requirement, as well as all of.the Baddam Neck technical specifications, will be reevaluated under SEP Topic XVI,
" Technical Specifications."
(c)
The fluid discharge through the RHR re).ief valve goes to the RWST and is still available for RCS cooling via the high and low pressure safety injection systems. However, during post-LOCA recirculation, the fluid may be radioactively centaminated, and leakage of this fluid to the RWST would not be ptable. One isciation valve is installed in the RER relief piping at the lef valve inlet and another in the relief valve discharge line outdoors near the RWST.
These v.tives are both manually operated. The isolat cn valve 4
n, ear the RWST ( SI-V-67 8) can be shut to isolate RER recirculation flow leakage dhrough the RER relief valve.
Cenclusion:
The RER relief valves meet the criteria of ETP 5-1 concer,ning discharge and collection of liquid from a stuck-open relief valve.
Fecuirement:
- C.3 If interlocks.are provided to automatically close the isolation valves when the RCS pressure exceeds the RER system design pressure, adequate relief capacity shall be l
provided during the time period while the valves are closing."
Evaluation:
- As noted in Section 4.1, these interlocks are not provided. However, as noted in Section 4.2, overpressure protection is afforded by the RHR ' relief valve in conjunction with the CPS, which combine to provide adequate relief capacity to prevent RCS pressure from exceeding RER design pressure.
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4 B-68 l'9.9 Frankfin Research Center A Demon of The Franulen Weehne l
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==
Conclusion:==
The lack of automatic isolation valve closure is acceptable because the combined capacity of the RHR and OPS relief valves meets pressure relief requirements.
4.3 PUMP PRCIECTION FIQUIRE.wENTS Recuirement:
"D.1 The design and cperating procedures of any RER system shall have provisions to prevent damage to the RER system pu=ps due to overheating, cavitation or 1 css of adequate suction fluid."
Evaluation:
.he features of the Haddam Neck plant designed to prevent damage to the
. pumps are provisions for pump cooling, a flow recirculation line, and a low flow alarm. Also, indications are available in the control room for RHR flow and valve pcsitions of all remotely operated RER valves. Either the service water system or the ccmponent cooling water system can be aligned to prov$de cooling water to the RER pump bearings and lubricating oil cooler to help prevent pump overheating. An alarm is provided to alert the operator to a hign temperature condition in the pump bearings.
In addition, a 3/4-inch line permits the recirculation of some RER pump flow from the discharge side to the j
suction side of the pump to prevent the absence of flow from overheating the I
operatir., pump. An RHR system low flow alarm alerts the operator if a low flew condition occurs while an RER pump is in operation. The availability of adequate net positive suction heat (NPSH) will be evaluated during the SEP r'eview of Topic VI
'.E, "ECCS Sump Design and Test for Recirculation Mode Effectiveness."
i Cenelusion:
The design and operating procedures of the RHR system are adequate to prevent damage to the RER system pumps in accordance with BTP 5-1.
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TER-C5257-311 4.4 TEST FIQUIRE.*ENTS Recuirement:
"E.1 The isolation valve operability and interlock circuits must be designed so as to permit on line testing when operacing in the RER mode. Testability shall meet the requirements of IEEE Standard 338 and Regulatory Guide 1.22."
i Evaluation:
RER isolatien valves cannot be tested during RER system operation'because valve closure will isolate the RER system from the RCS, Since the system lacks the type of required interlock that shuts an open valve if RCS pressure rises above the RER system design pressure, testing of such an interloct is not applicable.
(The staff concluded previously in Section 4 that lack of th,is type of interlock is acceptable in this case because of adequate empensation by procedural and administrative valve controls and the 400-psig CPS alarm.)
On-line testing does not apply to the existing interlock, which preve nts valve opening unless RCS pressure 'is less than the RER system design pressure, because it performs its function befera RER cooling, starts.
gnelusion:
The required test of the operability and interlocks of RER isolation valves is not applicable due to adequate administrative, procedural, and operational (i.e., CPS) valve controls.
1 I
Recuirement:
"We preeperational and initial startup test prograin shall be in conformance with Regulatory Guide 1.68.
The programs for PWRs shall include tests with supporting analysis to confirn (a) that adequate mixing of borated water added prio; to or during cooldown can be achieved under natural circulation conditions and permit estimation of the times required to achieve such mixing and (b) that the cooldown under natural circulation conditions can be achieved within the limits specified in the emergency operating procedures. Comparisen with performance of previously tested plants of similar design may be substituted for these tests."
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TER-CS 257-311 Evaluation:
Regulatory Guide 1.68 was not in existence when the Haddam Neck preoperational ahd irttial startup testing was accomplished. However, a test was perfcrmed to confirm that cooldown is possible by natural circulation [15].
The test involve,d timing the transit of a cold " slug" of RCS water driven by n a t.4. circulation around one of Haddam Neck's four RCS coolant locys.
Results indicated that natural circulation can attain a loop flow of approxi-mately 3% of design loop flow about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> af ter reactor shutdewn. This test and subsequent observations of natural circulation demonstrate adequate flow for core cooling.
No testing has been performed at Haddam Neck to determine the adequacy of bcron mixing under natural circulation flow conditions. However, the staf f f believes that, with the boric acid concentrations used for shutdown, natural circulation will produce adequate boron mixing.
Conclusien:
Precperational and initial startup test programs are not required, since Haddam. Neck began operating before Regulatory Guide 1.68 was issued.
BTP RSB 5-1 requirements for testing and analysis of natural circulation i
flow have been satisfied in Reference 15.
Analysis of boric acid concentratiens used for chutdown indicate that natural circulation will produce adequate bcron mixing.
4.5 OPERATIONAL PROCEDURES Requirement:
o
" F.1 The operational procedures for bringing the plart from normal operating power to cold shutdown shall be in '
conformance with Regulatory Guide 1.33.
For pressurized water reactors, the operational procedures shall include specific procedures and informatior. required for cooldewn under natural circulation conditions."
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TER-C5257-311 Evaluation:
Operaticnal procedures reviewed in this comparison of Haddam* Neck to BTP RSL 5-1 are discussed in Sect.on 2.
All of the procedures require the use of non-safety-crade equipment for portions of the shutdown operation. No procedures exist for shutdown and cooldown using safety-grade equipment only, but procedures do exist for achieving cold shutdown from outside the control room and for cooldown using natural circulation. The need for shutdown and cooldown procedures using only safety-c.ie equipment is not exp:essed in Regulatory Guide 1.33 but stems from BTP RSB 5-1 and SEP Topic VII-3.
The staff will consider requiring the Licensee to develop these procedures daring the integrated SEP assessment of the plant.
Conclusion:
- ' The procedures for safe shutdown and cocidown conform with Regulatory i
i Guide 1.33.
The staff will consider requiring the Licensee to develop procedures for shutdown and cooldown using only systemt on the minimum systems list during the integrated SEP assessment of the plant.
4.6 ACXILIARY FEEDWATER SUPPLY Recuirement:
"G.1 The seismic Ca,tegory I water supply for the auxiliary feedwater system for a PWR shall have sufficient inventory to permit operation at hot shutdown for at least four hours, followed by cooldown to the conditions l
permitting operation of the RHR system. The inventory needed for cocidown shall de based on the longest cooldewn time needed with either only onsite or only offsite power available with an assumed single failure."
Evaluation and
Conclusion:
See Section 3.2 for the discussion tnd conclusions regarding this BTP provision.
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TER-C5257-311' 5.
RESCLUTICN OF SEP "t) PICS The SEP topics associated with the saf e shutdown have been identified in 4
the introduction to this assessnent. The followir4g is a discussion of how the Haddam Neck plant meets the safety objectives of 'htsse topics.
5.1 Teoic V-10.3 RER System Reliability The saf ety objective of this topic is to ensure reliable plant shutdown
?
capability using safety-grade equi;mnt subject to the guidelines of SRP 5.4.7 i
The Haddam Neck systems have been compared with these criteria, and the results of these ecmp,arisons are discussed in Sections 3 t ad 4 of this assessmen:. Based en these discussions, the staff has concluded l that the systems fulfill the topic safety objective subject to resolution in the SEP integrated assessment cf the requirement fcr plant operating i
peccedures to shut down and cool dcwn using only systems identified in the I
minimum systems list.
I 5. 2 ~ ~to ic V-ll. A Recuirements fer Isolatien cf Hign and Lew Pressure Systems The saf ety objective of this topic is to assure that adequate measures are taken to protect the icw pressure system connected to the primary system i
frem being subjected to excessive pressure that could cause failures and in scxne cases pctentially cause a LOCA outside of contaiment.
This topic is assessed with regard to the isolation requirements ef the RER system frem the RCS.
As discussed in Sections 4.1 and 4.2, adequate overpressure protection exists for the RER system, subject to the following items, which will be considered in the SEP integrated assessment:
1.
the need for technical spe::ifications to require placing the overpressure protection system in ops. ration whenever RER cooling is in progress l
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TER-C5257-311 2.
the need for in' r.rlocks on the RER-to-core deluge motor-operated salves to prevent opening until RCS pressure is below RER design pressure 3.
the need for independent diverse interlocks on the RER-to-RCS isolation valves to prevent opening until ROS pressure is below RER design pressure.
5.1 Topic V-ll.B RRR Interlock Recuirements The safety objective of this topic is identical to that of Topic V-11.A.
The staff conclusion regarding the Ba ddam Neck v'4/e inter Accks is discussed in Sections 4.1 and 5.2.
In addition to these requirements, and as a matter to be resolved separately from the SEP, the NRC staf f has determined that certain isolation vsive configurations in systems connecting the high-pressure primary coolant system (PCS) to lower-pressure systems extending outside containment are a'
potentially significant contributors to an intersystem LOCA.
Such ccnfigurations have been found to represent a significent factor in the risk computed for 'ecre me,1t accidents (WASH-14 00, Event V). The sequence of events leading to the cere melt is initiated by the failure of two in-series check valves to function as a pressure isolation barrier between the high-pressure l
PCS and a lower-pressure system extending beyond containment. This causes an i
~
overpressurization and rupture of the low-pressure system, which results in a LOCA that bypasses c:ntainment.
The NRC has determined that the probability of failure of these check valves as a pressure isolation barrier can be significantly reduced if the pressure at each valve is continuously monitored or if each valve is periodically inspected by leak testing, ultrasonic examination, or radiographic inspection. The NRC has establish'ed a program to provide increased assurance thst such multiple isolation barriers are in place in all operating light water reactor plants. This program has been designated Multiplant Action Item MPB-4 5.
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TER-C5257-311 The Haddam Neck plant has modified its Technical Specifications to incorporate periodic leak testing for check valves in the RER-to-core deluge lines and in the high pressure safety injection cold leg lines.
5.4 Toeic VII-3 Svste:as Recuired for Safe Shutdown The safety ebjectives of this topic are:
1.
to assure the design adequacy of the safe shutdown system to:
(a) automatically initiate the operation of appropriate systems, including the reactivity control systems, so that spec.fied acceptable fuel design limits are not exceeded as e result of anticipated operational occurrences or postulated accidents, and (b) initiate the operation of syste.as and components recuired to bring the plant to a safe shutdown
.h to assure Ebat the required systems and equipment, 2.
including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, are in aporopriate locations outside the control room and have the capacity to subsequently bring the reactor to cold shutdown through the use of suitable procedures 3.
to assure that cnly safety-grade equipment is required for a ?WR plant to bring the reacter coolant system from a high pressure condition to a icw pressure cooling condition.
Safety objective 1(a) will be resclved in the SEP Design sasis Event reviews. Thase reviews will determine the acceptability of the plant including automatic initiation of safe-shutdown-related systems, to
- response, various Design Basis Events., i.e., accidents and transients [11].
Objective 1(b) relates to the availability in the centrol room of the l
. control and instrumentation systems needed to initiate the safe shutdown systems and assures that the control and instrumentation systems in the l
control room are capable of fellowing the plant shutdcwn f rom its initiation to its conclusion at cold shutdown conditions. The ability of the Haddam Neck plant to fulfill objective 1(b) is discussed in the ~ preceding sections of this Based on these discussiens, the staff concludes that safety objective report.
{
1(b) is met by the safe shutdown system at Haddam Neck, subject to the findings of related SEP Electrical, Instrumentation, and Control Topic reviews.
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TER-CS257-311 Saf ety objective 2 requires the capability to achieve both hot shutdown and cold shutdown conditions using systems instrumentation and controls located outside the control room. The Haddam Neck procedure AOP 3".2-8,
" Plant Operation Outside Control Room," provides the necessary steps to take the plant to the het shutdown condition and to proceed from there to the cold shutdown condition. Tne procedure contains the detciled steps required to achieve the coif shutdown condition, but it does not define the water sources for the auxiliary feedwater pumps. In response to Bulletins and Orders Task Force concerns, the Licensee hcs developed procedures for transf er of auxiliary feedwater suction to the alternate water supply. Portable battery-powered instrumentation is provided at the emergency control point in the cable vault penetration area to measure pressurizer pressure and level, reactor coolant temperature, and steam generator level; however, the procedure does not contain a program to ensure that the batteries are functional. The various work locations where operator action or attendance is required have been described in the procedure; however, the duty stations of the individual operators have not been defined. The procedure does not address the need for emergency communication equipment at the various duty stations.
Local instrumentation is used for the boric acid mix tank level, service water flow, and steam generator pressure.
The emergency control point is in the lower level'ef the cable vault.
The cable vault area has.an automatically initiated CO fire pr tection 2
system.
It appears that the initiation of the CO2 "Y*t** **I r*E*i '
evacuation of the emergency control point. In this case, the time period during which the control point would be unattended would be very short and would have a negligible icpact on the shutdown and cooldown procedure.
Based on the information provided in Procedure AOP 3.2-8 and obtained during the safe shutdown site visit, the staf f concludes that the Haddam Neck plant meets safety objective 2 of Topic VII-3, with the exception of procedural shortcomings regarding maintenance of batteries for portable instruments, assignments of shutdown duties for shif t personnel and emergency communication methods. The Licensee will be requested to modify procedures to alleviate
' he se shortcomings.
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TER-C5257-311 Whether the safe shutdown systems at Haddam Neck are safety-grade in conformance with safety objective 3 will be judced in part under SEP Topic III-1, " Classification of Structures, Components, and Systems (Seismic and Quality)," and in part under the Design Basis Event reviews. Table 3.'l of this report will be used as input to Topic III-1.
5.5 Tocic X Auxiliarv Feed System (AFS)
The safety objective of this topic is to assure that the AFS can provide adequate cooling water for decay heat removal in the event of loss of all main feedwater, using the guidelines of SRP 10.4.9 and BTP ASB 10-1.
The Haddam Neck AFS is described in rection 3.2.
This system has been compared with SRP 10.4.9 and STP ASB 10-1 with the following cenclus. ions:
f.
1.
The Haddam Neck plant, including the AFS, will be reevaluated during the SEP with regard to internally and externally generated missiles, pipe whip and jet impingement, quality and seismic design requirements, earthquakes, tornadoes, floods, and the failure of nonessential systems.
2.
The AFS conf rems to Genercl Design Criteria (GDC) 19,
" Control Room"; GDC 45, " Inspection of Cooling Water System"; GDC 46, "Testi.;g of Cooling Water Systems"; and Regulatory Guide 1.62, " Manual Initiation of Protection Actions." General Design Criterien 5, " Sharing of 7
Structures, Systems, and Components," is not applicaole.
3.
A passive failure of the ecmmon pump suction header, discharge header, or the non-essential condensate service line, to which the AFS suction line is attached, would prevent the AFS f rom scoplying feedwater to the steam generators even without an assumed concurrent single active failure. The low probability of a passive failure in the low pressure suction line or in the discharge line, which is periodically tested under the Licensee's inservice inspection program, alleviates the need for any immediate corrective measures. The staff intends to examine the need for a long-term improvement in the redundancy of the AFS at the Haddam Neck plant.
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TER-C5257-311
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4.
Although the AFS does not meet the provision for power diversity of BU ASB 10-1, the system design does permit emergency feeding of the steam generators with an assumed loss of all ac power, but manual operation of valves in the steam supply lines to the AFS turbines is required. In this case, manual valve operation is permissible because the steam generator water inventory can remove decay heat for approximately 40 minuten, with no feed.
5.
The staff is continuing to evaluate feed system waterhammer for the Eaddam Neck plant on a generic basis. SEP Topic V-13, "Waterhammer," applies.
6.
The AFS is not automatically initiated and the design does not have the capability to automatically terminate feedwater flow to a depressurized steam generator and provide flow to the intact steam generator. This is accomplished by the control room operator. The effect of this provision will be assessed in the main steam lir.e break evaluation for Haddam Neck.
7.
The technical specifications for the AFS will be j
reevaluated against current requirements under SEP Topic XVI, " Technical Specifications."
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TER-C5257-311 6.
REFERENCES 1.
" Report on the Systematic Evaluation of Operating Facilities" NRC Office of Nuclear Reactor Regulation November 25, 1977 2.
" Systematic Evaluation Program, Status Summary Report" NRC 3.
" Staff Discussion of Fifte'en Technical Issues" In attachment to November 3, 1976 memorandum from Director, NRR, to NRR Staff Nove.2er 1976 4.
Facility Description and Safety Analysis Report for the Haddam Neck Plant as amended 5.
10CFR50, Appendix B Supplement to the Safety Evaluation by the Directorate of Reactor Licensing USAEC December 27, 1974 6.
Supporting Information
',r the Connecticut Yankee Full Term Operatine License Application December 1969 f
7.
W. G. Counsil (CYAFCO)
Letter to D. L. Ziemann (NRC)
Subject:
Remarks on SEP Draf t Assessment September 28, 1979 8.
D. Switzer (CYAPCO)
Letter to A. Schwencer (NRC)
Radcam Neck Inservice Inspection Program June 29, 1977 9.
W. G. Counsil (CYAPCO)
Letter to H. R. Denton (NRC)
December 31, 1979 10.
D. Switzer (CYAPCO)
Letter to A. Schwencer (IGC)
March 1, 1977 B -7 9
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TER-C5257-311 e
11.
Specific Plant Report
" Low Temperature RCS Overpressure Protection for Connecticut Yankee"
( Aue ::st. 1977)
, to letter from D. Switzer (CYA"CO) to A. Schwencer (NRC)
Feptember 7, 1977 12.
" Pressure Mitigating System Transient Analysis Results" prepared by Westinchouse for the W Users Grcup on teactor Coolant Syatem Overpressuri:ation, July, 1977 to letter from D. Switzer (CYAPCO) to A Schwencer (NRC)
September 7, 1977 13.
D. L. Zieman (NkC)
Letter to W. G. Counsil (CYA7'O
Subject:
Amendment 33 to Lietnt.c April 24, 1980 14.
Technical Specifications Appendix A to Facility Operating License DPR-61 for the Haddam Ne2k Plant 15.
CYAPCO October 1969 Report, " Reactor and Plant Performance Engineering Tests and Maintenance" As appended tv ALC Division of Compliance Letter Oecember 1969 16.
W. G. Counsil (CYAPCo)
Letter to D. M. Crutchfield (NRC)
July 22, 1981 A
B -8 0 i $'. Franklin Research Ccater L
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TER-C5257-311 APPENDIX B, PART 2
(
SAFE SHUTDCWN WATER REQUIRE. VENTS Introduction Standard Review Plan (SRP) 5 4.7, " Residual Heat Removal (RHRI System";
Regulatory Guide 1.139, " Guidance for Residual Heat Removal"; and Branch
'Nchnical Position (BTP) RSB 5-1, Rev. 1, " Design Requirements of the Residual Heat Remeval System," are the current criteria used in the Systematic Evaluation Procram (SEP) evaluation of systems required for safe shutdown.
BTP RSB 5-1 Section A.4 states that the safe shutdown system shall be capable of bringing the reactor to a cold shutdown condition, with only cffsite or onsite power aveilable, within a reasonable period of time following shutdown, assuming the most limiting single failure. BTP RSB 5-1 Section G, which i!. applies specifically to the amount of auJiliary feed system (AFS) water of a pressuri:ed water reactor available for cteam generator feeding, reqcires the seismic Category I water supply for the AFS to have sufficient inventory to permit cperation at hot shutdown for 'at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, fc11 owed by cooldown to
~
u 3 conditions permitting operation of the RER system. The inventory needed ror cocidown shall be based on the longest ecoldewn time needed sith either only ensite or only offsite pcwer available with an assumed single failure.
A reasonable period of time to achieve cold shutdown conditions, as stated in SRP 5.4.7 Section III.5, is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
For a reactor plant cooldown, water is the medium for transfer of heat from the plant to the a nvirons. Two modes of heat removal are available. The j
first mode involves the use of reactor plant heat to boil water and the venting of the resulting steam to the atmosphere. The water for this process
~ is typically demineralized " pure" water stored onsite and, therefore, is limited in quantity. The systems designed to use this mode of heat removal (boilof f) are the steam generators for a pressurized water reactor (FWR). The second heat removal mode (blowdown) involves the use of power-cperated relief valves to remove hea't in the form of steam energy directly from the tsactor coolant system. Since it is not acceptable to vent the reactor coolant system A@) Frankin Reser.h Center B -81 A.
A h of he Frecu tasanne
TER-CS257-311 directly to the atmosphere, the steam is typically vented to the containment building from which containment cooling water systems transfer the heat to an ultimate heat sink, usually a river, lake, or ocean. When the blowdown mode is used, reactor coc hnt system makeup water must be continuously supplied to keep the reactor core covered with coolant to compensate for loss of co'lant o
inventory. The efficacy of the blowdown mode for PWRs has received increased staff attention since the Three Mile Island Unit 2 accident in March 1979.
Additional studies are in progress or planned.
This evaluation of cooling water crquirements for safe shutdown and cooldown is based on the use of the system identified in the SEP Review of Safe Shutdown Systems which has been completed for each SEP facility in accordance with SRP 5.4.7 and BTP RSB 3-1 criteria.
It should be noted that the SEP Design Basis Events (DBE) reviews, now in progress, may require the ~
use of systems other than those evaluated in this report for reactor plant
.kNutdownandcooldown.
In those cases, the water requirements for safe shutdown will have to be evaluated using the assumptions of the DBE review.
1 Discussion The requirement in BTP RSB S-1 and SRP 5.4.7 that a plant achieve a cold ihutdcwn condition within approximately 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> is based mainly on the desire to activate the RFR system and transfer the plant heat to an ultimate heat sink prior to the exhaustion of the limited amount of onsite-stored pure water available for the AFS of a PWR.
A sustained hot shutdown condition, with reactor coolant systems temperature and pressure in excess of RHR initiation limits, requires continued boiling off of pure water to remove reactor core decay heat.
If onsite-t'.ored water at a plant is depleted, raw water from a river, lake, or ocean can usually be tapped to supply the boiloff systems. However, raw water can accelerate the corrosion of boiloff system materials in the steam generator and energency condenser tubes even if the water is fresh.
Seawater can cause chloride stress-corrosion cracking of the tubes within one week. Raw fresh water can cause caustic stress-corros'.on cracking of both A
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TER-CS257-311 stainless steel and inconel tubes in less than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> through NaOH concentration. Plant cooldewn and depressurization would help reduce the rato of tube cracking by reducing the stresses in the tube materials, and.would also reduce the leakage race of reactor coolant through cracks that do occur.
The original design criteria for the SEP facilities did not require the ability to achieve cold shutdown conditions. For these plants, and for the majority of operating plants, safe shutdown was defined as hot shutdown.
.Therefore, the design of the systems used to achieve a cold shutdown cQnd,iti,on
' ' ~
was determined by the reacter plant vendor and was not necessarily based on safety concerns. Safe shutdown reviews have pointed out a difference in vendor approaches to system design for cold shutdown reflected in the Standard Tech-
.nical Specification definitica of cold shutdown. For a BWR, cold shutdown for a PWR, the temperature l requires reactor coolant temperature to be < 212*F; is < 200*F.
This difference in cold shutdown temperatures requires additional systems for PWR cooling not needed for a BWR.
For example, a BWR could use an isolation condenser alone to reach 212*F (although the approach to the final temperature would be asymptotic), but a PWR, in addition to the st,eam generators, must use an RHR and supporting systems to cool to 200*F.
Evaluation Table B.1-1 provides plant-specific data and assumptions used in the staff calculation of safe shutdown water requirements for the Raddam Neck Unit 1 plant. Table B.2-2 presents the results of the calculation.
Upon a less of lead, steam released to the atmosphere and condenser will limit the reactor coolant system pressure and temperature transient. If the
. load decrease exceeds the capability of the atmospheric dump valves and steam I
bypass valves, a reactor trip will be initiated to protect the core.
The l
atmospheric dump valve and the steam re merator vent system are the primary I
(
decay heat removal mechanisms available initially to begin a cooldown af ter heat removal capacity matches decay heat input.
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$LEnk:in Research Center A Dwsen af The Franeniasstwo i
TER-C5257-311 The decay heat release header (DHRH) upstream of the main steam non-return valves is t?e source of steam for the ADV, steam generator vents, and AFP turbines, while the hogging jets and SJAEs are supplied from downstream of the NRVs.
Steam for the DERH is provided from all four steam gene rtors. A single failure of the most limiting component (a steam generator vont) following disablement of the ADV due to loss of control air will not prevent the plant from maintaining hot shutdown or from cooling down.
Section 3.2 indicates that the heat removal rate of these steam generator vents and the auxiliary vent paths equals the decay heat input rate approxi=ately 3.3 minutes after reactor scram. Since sufficient steam removal capacity is available to cool down in approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with loss of the ADV and a single steam generator vent, a single failure in the decay heat removal system is not the most limiting influence on cooldown rate. There-fore, the s m iliary feedwater system was examined to evaluate its effect on cocidown time and minimum water requirements.
ROS pressure is maintain $d near normal operating pressures by venting the system to the atmosphne for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to commencing the i
cooldown. The 4-hour celay is based on BTP RSB 5-1 Section G and again is intended to maximize pure-water consumption, 6
Four hours after reactor trip, the decay heat rate is 53.6 x 10 Etu/h i
0 and the integrated heat ove: the 4-hour period is 3.02 x 10 Btu.
Assuming that the plant operator.has maintained a constant mass of water in the steam generator and has maintained steam generater pressure by throttling the ADVs, l
32,400 gallons of auxiliary feedwater is required to remove the integrated heat.
Scoping calculations indicate that if auxiliary feedwater flow is initiated at design flow after 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the cooldown rate from a single AFP would exceed 225'F/h. This cooldown rate exceeds the admin 2.trative limit of Sb'F/haswellastheTechnicalSpecificationslimitof100*F/h. To reduce the cooldown to the administrative limit while maintaining a constant steam generator level, the operator must throttle auxiliary feedwater flow as well as the steam flow to the atmosphere. These calculations also demonstrate that the 4-hour delay prior to st uting cooldown permits the decay heat rate to e ~s E-8 4 P
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decrease sufficiently so that a single failure impairing auxiliary feedwater flow is not significant in extending the cooldown time.
Scoping calculations assuming a constant cooldown rate at the administrative limit indicate that,the cooldown rate of 50*F/h cannot be maintained throughout the cooldown to 350*F.
The quantity of demineralized water consumed at the end of the 4-hour delay period is 32,400 gallons.
Cooldown to the conditions permitting RER system operation regires an additional 19.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and consumes 97,600 gallons. Although the DWST has a capacity of 100,000 gallons, Technical Specification.3.8.A requires a minimum of only 50,000 gallons, in addition to at least 80,000 gallons from the PWST.
Based on these calculations, sufficient makeup inventory capacity is available E.o conduct a plant cooldewn in accordance with BTP RSB 5-1.
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Table B.2-1.
PLAN *-SPECIFIC DATA AND COOLDOWN ASSUMPTIONS Plant-Soecific Data Haddam Neck Plant Power 1825 MW
~
Initial ROS Temperature 544*F Reat Removal Capacity
- 1.624 x 108 Etu/h Auxiliary Feedwater Pump Capacity 450 gpm (turbine-driven)
Secondary Makeup Water Temperature 100*F Stored Sensible Heat Metal - d.89 x 105 Btu /*F Water - 4.14 x 105 Btu /*F
- t.
Total Feedwater Available 130,000 gal Coeldown Assumotions 1.
Reactor trips at t = 0.
2.
Decay heat is in accordance with Draft ANS-5.1.
3.
Plant remains at het shutdown for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to cooldown.
4.
Mwss of water d.n the steam generators is constant.
5.
Feedwater and steam flowrates are throttled to maintain the administrative cooldown rate.
- Using 3 steam generator vents and auxiliary vent paths.
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TER-C5257-311 s
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Table B.2-2.
CALCULATION OF SAFE SH'G CWN WATEn REQUIRE 5INTS b
Plants Haddam Neck Phase I Reactor trip to point at which decay heat generatlon rate equals heat removal capacity: approx. 3.3 min
- Phase II Delay prior to cooldown: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Decay heat generated prior to cooldown: 3.02 x 10 Btu Feedwater expended prior to cooldown:
32,400 gal Phase III Cooldown*:
Steam RCS Generator Time (h)
Tercerature (*F)
Pressure (psia)
Decav Heat Generated (Btu) 4 544 1,000 3.02 x 10 6.9 467 500 4.43 x 10 8
23.6 350 149.3 1.07 x 10 Decay heat rate at t = 23.6 hs 31.1 x 106 Btu /h Feedwater expended during cooldown to 350*F: 97,600 gal Total feedwater expended: 130,000 gal
- Using 3 steam generator vents and auxiliary vent paths.
B-87 MJ Frankfin Research Center A No.oa e n. r.
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1