Text

Forwards Revised SEP Review of Safe Shutdown Sys for Yankee Rowe Nuclear Power Plant
	ML20005B896
Person / Time
Site:	Yankee Rowe
Issue date:	09/06/1981
From:	Crutchfield D; Office of Nuclear Reactor Regulation
To:	Kay J; YANKEE ATOMIC ELECTRIC CO.
References
	TASK-05-10.B, TASK-05-11.B, TASK-07-03, TASK-5-10.B, TASK-5-11.B, TASK-7-3, TASK-RR LSO5-81-09-020, LSO5-81-9-20, NUDOCS 8109160087
	Download: ML20005B896 (92)
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U September 6,1981 6

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!!r. James A. Kay

, %=mmans Senior Engineer - Licensing 8

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Dear !!r. Kay:

SUBJECT:

YNil'IE ROUE - SEP TOPICS V-10.B. RilR RELIABILITY, V-11.B.

RilR IllTERLOCK REQUIRE!!EllTS, NiD VIb3, SYSTDIS REQUIRED FOR SAFE SliUTDOUN (SAFE SilUTDOUll SYSTEllS REPORT)

The SEP draf t ctaluation of Safe Shutdown Systens for Yankee Rowe was transmitted to you on. ebruary 14, 1979.

iour cor.rier,ts were provided in a letter dated April 13, 1979.

The draft Safe Shutdown Systems Report has been revised by contractor personnel to resolve your cor.rients, to establish the minimum set of systems 1

needed to satisfy the staff's functional criteria and to detemine the water required to perfom the cooldown in accordance with these criteria.

The revised eval d tfon is enclosed. You are requested to infom us if your as-built facility differs from the licensing basis assumed in the assessment within 30 days of receipt of this letter.

This topic evaluation will be a basic input to the integrated safety assess-4 ment unless you identify changes needed to reflect the as-built conditions i

of your facility. This topic assessment may be revised in the future if your facility design is changed or if !!RC criteria relating to this topic are codified before the integrated assessmnt is completed.

Also enclosed are the conclusions the staff has fomulated as a result of these topic evaluations and identification of concerns.for resolution in the integrated ascessment.

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2 You are encouraged to supply for the docket any other material related to these coi.cerns that may affect the staff's evaluation and conclusions, to enable us to resolve these deviations prior to the integrated assessment.

Sincerely, Dennis H. Crutchfield, Chief Operating Reactors Branch fio. 5 Division of licensing

Enclosures:

Safe Shutdown Systems Report and Staff Conclusions cc w/ enclosures:

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o Mr. Ja:res A. Kay cc Er. James E. Tribble, President Ydnkee Atcmic Electric Company 25 Research Drive Westborough, Massachusetts 01581 Greenfield Community College 1 College Drive Greenfield, Massachusetts 01301 Chairman Board of Selectmen Town of Rowe Rowe, Massachusetts 01267 Energy Facilities Sitir4, Council 14th Floor One Ashburton Place Boston, Massachusetts 02108 U. S. Environmenta*. Protection Agency Region I Office ATTN: EIS COORDINATOR JFK Federal Bu lding i

Boston, Massachusetts 02203 Resident Inspector Yankee Rowe Nuclear Power Station c/o U.S. NRC Post Office Box 28 Monroe Bridg;, Massachusetts 01350 l

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e CONCLUSIONS 1.

The staff concludes that the current

steam relieving paths do not conform to Criterion 4 and that proposed modifications (installed during Cycle XV reload) as described would provide sufficient capacity and redundancy to satisfy the functional requirements of Branch Technical Position (BTP)

RSB 5-1.

2.

The staff concludes that the current *:uxiliary feedwater system doet not meet the functional requirements of BTP RSB 5-1 but that proposed modifica-tions as described would satisfy the functional requirements of BYP RSB 5-1.

3.

The staff concludes that the shutdown cooling system '3CS), the component cooling water system (CCWS) and service water system (SWS) satisfy the functional requirements of BTP RSB 5-1, except that the electrical com-ponents are not powered from diesel-supplied electrical buses The staff will evaluate the significance of this in the SEP integrated assessment of Yankee Rowe.

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The staff defers evaluation of the adequacy of the pressum control and relief system to satisfy BTP RSB 5-1 pending resolution of current staff reviews of applicable TMI-2 action items 2nd fire protection requirements.

't he staff will determine the effect that the completion of these reviews will have on the safe shutdown topic during the integrated assessment.

5.

The staff concludes that the chemical volume and control system does not meet the functional requirements of BTP RSB 5-1 in that the charging and letdown paths are susceptible to single failure and that the charging pump and other electrical components are not powered from diesel-supplied electrical buses.

The staff will avaluate the sigr'ficance of these devia-tions during the SEP integrated assessment of Yankee Rowe.

6.

The staff concludes that the control air system does not satisfy the functional requirements of BTP RSB 5-1 in that a reliable source of control air is not available and significant operator action outside the control l

room would therefore be required to effect a safe shutdown.

The staff will evaluate the significance of this in the SEP integrated assessment i

of Yankee Rcwe.

Carrent denotes pre Core XV configuration.

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The amount of operator action required to perform the cooldown to cold shutdown is not etmpatible with the intent of the topic criteria.

The staff will consider the need for increased control room operability of cooldown systems during the integrated assessment.

8.

Due to the potential severity of SCS overpressurization, the steff will consider requiring the following during the integrated assessment:

(1) interlocks to prevent opening of SCS isolation valves until the main coolant system pressure is below SCS design pressure and I

/C) valve position indication for the isolation valves in the control room.

The staff has determined that the installation of automatic closure Interlocks would.not be desirable sirice two of the three low temperature overpressure protection (LTOP) relief valves are on the SCS, and automatic isolation of the SCS from the reactor coolant system (RCS) would render the LTOP system inoperable.

However, in the SEP integrated assessment the staff will evaluate the potential need for additional measures, such as control room valve indications, to prevent RCS startup and pressurization with any SCd isolation valves ir the open position.

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SEP Review of Safe Shutdown Systems for the I

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TER-<5257-310 CONTENTS Section P. ale, 1

INTRODUCTION.

B-1 2

DISCUSSION B-6 2.1 Normal Pl?.nt Shutdown and Cooldown.

B-6 2.2 Shutdown and Cooldown with Loss of Cffsite Power.

B-8 3

CCNICP.vANCE WITH BRANCH TECHNICAL POSITICN 5-1 FUNCTICNAL REQUIRDiENTS.

B-10 3.1 Background.

B-10 3.2 Functienal Requirements.

B-19 l

3.3 Safe Shutdown Instrumentation B-56 4

SPECIFIC RESIDUAL HEAT REMOVAL AND OTHER RMUIRDiENTS OF BRANCH TIX'ENICAL POSITICN 5-1.

B-61 4.1 Pesidual Heat Removal System Isolation Recuirements.

B-61 4.2 Pressure Relief Paquirements B-63 4.3 Pump Protecticn Requirements B-69 4.4 Test Requirements B-70 4.5 Cperational Precedures.

B-71 4.6 Auxiliary Feedwater Supply.

B-71 I

5 RESCLUTION CF SYSTDiATIC EVALCATICN PROGRAM TCPICS B-73 1

5.1 Tcpic V-10.3 RER System P411 ability B-73 5.2 Topic V-ll.A Requirements for Isolation of High and 'Aw Pressure Systems and "opic V-ll.B RER Interlock Requirements B-75 l

5.3 Topic VII-3 Systems Required for Saf e Shutdown.

B-77 5.4 Topic X Auxiliary Feed System.

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TER-C5257-310 1.

INTRODUCTICN The Systematic Evaluation Program (SEP) review of the "saf e shutdown" subject encompassed all or purts of the following SEP topics, which are among

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those identified in the November 25,, 1977 NRC Office of Nuclear Reactor Regu-lation document entitled Report on the Systematic Evaluation of Operating Fac ilities":

1.

Residual Heat Removal System Reliability (Topic V-10.B) 2.

Requirements for Isolation of High and Low Pressure Systems (Topic V-ll. A) 3.

Residual Heat Removal Interlock Requirements (Topic V-ll.B) 4.

Systems Required for Safo Shutdown (Topic VII-3) 5.

Station Service and Cooling Water Systems (Topic IX-3) i 6.

Auxiliary Feedwater System (Tcpic X).

The review was primarily performed during an onsite visit by a team of SEP personnel. This onsite effort, which was performed from June 13 to June 16, 1978, af forded the team the opportunity to obtain current information and to examine the applicable equipment and procedures, and it also gave the Licensee (Yankee Atemic Electric Company} the opportunity to provide input into the review.

The review included specific system and wguipment requirements for remaining in a hot standby condition (defined as the reactor suberitical with T

reduced to between 530*F and 330*F) and for proceeding to a cold VE shutdewn condition (defined as reacter coolant temperature less than 200*F).

The review for transition from reactor operaticn to het standby considered the requirement for the espability to perform this operation from outside the control rocm. The review was augmented as neceasary to assure resolution of the applicable topics, except as noted below:

Topic V-11.A (Requirements for Isolation of High and Low Pressure Systems) was examined only for applicatien to the residual heat removal (RER) system. Cther high pressure / low pressure interfaces were not investigated.

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TER-CS257-310 Topic VII-3 (Systems Required for Safe Shutdown) was completed except for determinatien of design adequacy of t.e system.

Tepic IX-3 (Station Service and Cooling Water Systems) was caly reviewed to consider redundancy and seismic and quality classiffcatien cf cooling water systems that are vital to the performance of safe shutdown r.ystem components.

Topic X was reviewed only to address design adequacy for heat removal.

Other aspects are considered as part of the design basis event review cr under implementation of the ':MI Action Plan.

The criteria applied to the safe shutdown systems and ccmponents in this raview are taken from the Standard Review Plan (SRP) 5.4.7, "Fesidual Heat Femoval (RER) System"; Branch Technical Position RSB 5-1, Revisicn 1, " Design Requirements of the Pasidual Heat Removal System"; and Regulatcry Guide 1.139,

" Guidance for Residual Heat Removal." These documents represent current staff r.

criteria und are used in tr.a review of f acilities being processed for cperating licenses. This comparison of the existing systems with current licensing criteria led naturally to at least a partial consideration of design criteria that will be pertinent to SEP Topic III-1, "Classifi~ation of c

Structures, Components and Systems (Seismic and Quality)." This report will also be reviewd for its application to the resolution of other topics.

As noted above, tts six tcpics were examined while possible interactions with other tcpics, systems, and components not directly related to safe shutdown were neglected. For example, Tcpics II-3.3 (Flooding Pctential and Protecticn Requirements), II-1.C (Saf ety-Related Water Supply), III-4.C (Internally Generated Missiles), III-5.A (Effects of Pipe Break on Structures, Systems and Components Inside Containment), III-6 (Seismic Design Considerations), III-10.A (Thermal Overload Prctection for Motors of Motor-Operated Valves), III-ll (Compcnent Integrity), III-12 (Environmental Qualification of Saf ety-Related Equipment), and V-1 (Co:rpliance with Codes ud Standards) could be affected by the results of the safe shutdown review or could affect the safety of the systems that were reviewed. These effxts will be reviewed later. Further, this review did not cover in any significact A

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TER-C5 257-310 detail either the reactor protection cystem or the electrical power distribution, both of which will also be reviewed later.

The staff considers that the ultimate decision concerning the safety of any of the SEP fac'ilities is based upon the ability of the facility to withstand the SEP design basis events (DBEs). The SEP topics provide a major input to the DBE retiew, trom the standpoint of assessing both the probability and the consequences of the event. As examples, the safe shutdown topics pertaiwing to,the, listed DBEs are provided in Table B.1 (the extent of applicability will be determined during the plant-specific review).

Complation of the safe shutdown topic review (limited in scope as noted above), as docu: rented in the attached report, significantly contributes to an assessment of the existing safeq margins.

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TER-CS257-310 Table B.1 IMPACT UPON PRCBABILITY

'IVPIC DBE GRCUP CR CONSEQUENCES CF DBE V-10.B VII (Spectrum of Loss-of-Consequences coolant Accidents)

V-ll.A VII (Defined above)

Probability V-11.B VII (Defined above)

Probability VII-3 All (Defined as a generic topic)

Consequences IX-3 III (Stean Line Break Inside Consequences Containment)

(Steam Line Break Outside Concainment) l IV (Loss of AC Power to Station Consequences Auxiliary)

(Loss of all AC Power)

V (Loss of Forced Coolant Flow)

Probability (Primary Pump Rotor Seizure)

(Primary Pump Shaft Break) l VII (Defined above)

Consequences X

II (Loss of External Load)

Ccnsequences (Turbine Trip)

(Loss of Condenser Vacuum)

(Steam Pressure Regulator (closed])

(Loss of Feedwater Flow) l (Feedwater System Pipe Break)

III (Defined above)

Consequences IV (Defined above)

Consequences V

(Defined above)

Consequences VII (Defined above)

Consequences

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TER-CS257-310 Picina System Passive Failures The NRC staf f ncemally postulates piping system passive failures as (1) accident-initiating events in accordance with staff positions on piping failures inside and outside containment, (2) syste9 leaks during long-term coolant recirculatien following a ICCA, and (3) failures resulting from hazards such as earthq"skes and tornado missiles.

In this evaluation, certain l

piping system passive f ailures have been assumed beyond those normally postulated by d.e staff, e.g., the cat sstrophic f ailure of moderate energy sy stems. These assumptions were made to demonstrate safe shutdown system redundancy in the event of complete failures of these systems and to f acilitate future SEP reviews of LBEs and other topics that will use the safe shutdown evaluation as a source of data for the SEP facilities. SRP 5.4.7 and BTP RSB 5-1 do rot require the assumptions of piping system passive failures.

Credit for Operatinc Procedures For the safe shutdowri evaluation, the staff may give credit for facility operating procedures as alternate means of meeting regulatory-guidelines.

l Those procedural requirements identified as essential for acceptance of an SEP topic on DBE.3 will be carried through the review process and considered in the intecrated assessment of the facility. At that time, the staf f will decide I

which procedurus are so irportant that an administrative method must be estatif shed to ensure that, in the future, these operating procedures are not changed withcut appropriate connderation of their importance to the SEP topic evaluation.

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DISCUSSICN 2.1 Ncemal Plant Shutdown and Cooldown A normal.=butdown f cm full power to hot standby is accomplished with the use of operating procedure CP-2104,." Scheduled Plant Shutdown t0 Hot Standby." The shutdown from power is accomplished by reducing the generator load using the turbine control system and following with control rod insertion to control T The load 'r'edubt. ion.is performed at a rate of 8 MWe per 5 AVE.

minutes, and changes in main coolant average temperature are controlled at a rate of 2*F per 5 minutes. The reactor is borated using the charging pumps to the amount necessary to maintain the cetrol rod bank above the low insertion limit and ensure that the axial flux difference will remain within its target band.

l The first main feedwater pump and condensate pump are removed from service when the generator load has been reduced to less than 140 MWe.

When the generator load has been reduced to less than 60 MWe, the second main feedwater pump and condensate pump are removed from service. The power reduction is centinued uring the remaining operating boiler feed pump to pec4*ide feed to the stears generators. When the load on the generator has decreased to less than 30 MWe, station service loads are transferred to the auxiliary transformers fed frem the offsite power supply and ecndensate recirculation is established back to the condenser het well. Manual cortrol of the tud ne bypass is taken when the generator load is reduced to less than 15 MWe.

The turoine is tripped just before the generator lead reaches O MWe.

Normally, the plant can be maintained in a hot standby condition (main coolant average temperature at 514

F, 2000 psig) by using main coolant pump heat, decay heat, ar.I discharging steam to the main steam header.

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During the plant shutdown to hot standby, a centrol rod group remains withdrawn to a height sufficient to provide a reactivity worth of 1% for emergency shutdown capability.

If for any reason a centrol rod group cannot be withdrawn to provide a reactivity worth of 1%, then the main coolant system is borated to 5% ak/k shutdcwn margin. Prior to pecceeding to hot shutdown I

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o TER-C5257-310 and ~ cold shutdown, the main coolant system is borated to the 5% sk/k shutdown margin. The main coolant system is borated using the charging and volume control system. The charging pumps take suction from the boric acid mixing and storage tank. 'If any main coolant loops are isolated, they too are borated to the shutdown margin.

The plant cooldown is limited to 50'F per hour, and cooling is i

accomplished by continuing the bypass of steam to the main condenser. At le.ast two main coolant loops are tied to the reactor vessel until the shut-down coolang system is in operation. Pressurizer level is now manually controlled using the charging pumps to provide makeup for contraction caused by the cooldown of main coolant system water. Pressurizer temperature and pressure are controlled to maintain the -reactor vessel within nil ductility transition temperature range.

During the plant cooldown, one main coolant pump is opersted. When the main coolant temperature has been reduced to less than 330*F, pressure is adjusted and the shutdown cooling system is placed in operation. Af ter the shutdown cooling system has been placed in operaticit, the one remaining main coolant pump may be shut down.

If the main coolant system is to be depressurized, then the remaining main coolant pump is secured. Pressurizer temperature and pressure reduction is performed by charging pump flow through the auxiliary spray line to the pressurizer spray, or by pressurizer steam discharge through the soleroid relief valve to the low pressure surge tank.

When the main coolant syster. temperature reaches 200'F, the charging rate is increased to the main coolant system in order to fill the pressuri:er. The meter-operated relief valve is closed, the solenoid-operated relief valve is deenergized to close, and the pressurizer vent is opened to depressurize the main coolant system. Shutdown cooling continues until the main coolant system reaches about 14 0*F, where it is maintained by the shutdown coolir, system.

This system is cooled by the component cooling system, which is cooled by the service water system. The service water system takes cold water from the river, circulates it through the component cooling system heat exchangers and returns the warmer water to the river. Thus, heat is transferred from the main coolant system to the river to accomplish cocidown and decay heat removal.

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e TER-C5257-310 2.2 Shutdown and Cooldown with Loss of Of fsite Power Operating Procedure " Loss of A.C. Supply" defines the action to be taken following a total loss of ac power to provide emergency electrical power to vital equipment.

Following a loss of offsite power and turbine trip, the main condenser circulating water pumps cannot be powered from onsite sources. With the loss of circulating water pumps, main condenser vacuum cannot be maintained and the main condenser becomes unavailable fer heat removal. With the less of normal heat sink, the main steam safety valves will lif t to vent stear to atmosphere. The operator is directed to verify closing of the steam dump valve on loss of condenser vaccum and automatic starting of the three emergency power diesel generators. In addition, the operator is directed to perform the necessary electrical switching to remove connections to the c' fsite power lines and to start the emergency boiler feed pump. The steam f

supply valve to the large hogger is cpened and set to maintain an inlet steam 4

pressure of 300 psig. The emergency feed supply valves are cpened, and feeding to the steam generators is ccmmenced.

Electrical rower is restored to the 480-V buses, and pressurizer heaters Nos. 5 through 8 are energired to restore main coolant systea pressure control. The operator establishes a minimum of 200 psig overpressure on the main coolant system. A service water' pump and component cooling water pump are then started to supply plant equipment cooling requirements.

l CPerating Procedure " Loss of Condenser Vacuum" delineat'es the action te be taken if a loss of condenser vacuum occurs while the plant is eperating at power. One of the im;nediate actions is to initiate maximum feed and bleed and to, increase low pressure surge tank cooling, if required.

Subsequent cperater action is to line up the following equipment to provide main coolant heat removal to control temperature as necessary:

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atmospheric steam dump i

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steam drains to atmosphere d.

throttle line steam drains to the auxiliary boiler blowdewn tank.

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TER-C5257-310 At this point, the plant's essential equipment is being supplied through the operation of the emergcacy diesel generator. hactor decay heat is ttsnsferred to the steam generators and dissipated by lif ting of the main steam saf ety valves and operation of various other vent paths.

Operating Proceiure " Plant Cooldown from Hot Standby" delineates the 4

steps required for a plant cooldown with or without the main condenser in service. The operater is directed to do the following when the main condenser is not availt.ble:

borate the nin coolant system to the shutdown margin o

o initiate maximum feed with supplemental low pressure surge tank cooling o

adjust the atmospheric steam dump valve to achieve the desired cooldown rate but not greater than 50*F/h t

o remove the emergency core cooling system from service when the main coolant system pressure is less than 1000 psig o

initiate the shutdewn ecoling water system when the main coolant system pressure is less than 300 poig and temperature is between 300 and 330*F.

Cooling with the shutdown cooling system is then accomplished in the same manner as was discussed in Section 2.1.

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TER-CS 257-310 3.

CCNFORMANCE WITH BRANCH TECHNICAL POSITICN 5-1 FUNCTICNAL REQUIRC4ENTS The functional requirements stated in Branch Technical Position (DTP) 5-1 for the safe shutdown systems are:

1.

The design shall be such that the reactor can be taken frm.wrmal operating conditions to cold shutdown

using only safety-gride systems. These systems shall satisfy General Design Criterra 1 through 5.

2.

The system (s) shall have suit.tble redundancy in ecmponent.: and features, and suitable interconnections, leak detecticn, and isolatien capabilities to assure that fot onsite electrical power system cperation (assuming of fsite power is not available) and for of fsite electrical power system operation (assuming onsite power is not available) the system function can be accomplished assuming a single failure.

3.

The system (s) shall be capable of being operated frcm the control room with either only onsite or enly offsite power available with an assumed single failure.

In demenstrating that the system can per form its function assuming a single failure, limited operater action outside of the control room would be considered acceptable if suitably justified.

4.

The system (s) shall be capable of bringing the reactor to a cold shutdown condition,* with enly offsite er ensite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.

Compliance of the Yankee Powe safe shutdown systems with these criteria is discussed below.

3.1 Backcround The BTP 5-1 requirements are stated with respect to plant chutdown and cooldown with only offsite or only onsite power available. The staff evaluat'ed the plant's ability to cenduct a shutdewn with only offsite pmer available and determined that the "only onsite power available" case is more

" Processes involved in cooldewn are heat removal, depressurization, flow circulation, and reactivity control. The cold shutdown conditien, as described in the Standard Technical Specifications, refers to a suberitical reactor with a reactor coolant temperature no greater than 200'F.

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s TER-CS257-310 limiting. The plant electrical system is sufficiently versatile to allow energizing of all recessary equipment from only offsite power. Therefore, the staff concentrated its evaluation of the Yankee Bowe safe shutdown systems to shutdown following a loss of offsite power.

A "saf ety-grade" system is define 4, in the NUREG-0138 (1) discussien of issue No. 1, as one which 7.; designed to seismic Category I (Fagulatory Guide 1.29) Quality Group J or better (Regulatory Guide 1.26) and is operated by electrical instruments and contrc'.s that meet Institute of Electrical and Flectronics Engineers Criteria for Nuclear Power Plant Protection Systems (IEEE Std 279-1971). Yankee Powe received its Full Term Operating Licer.4e on June 23, 1961 prior to the issuance of Regulatory Guides 1.26 and 1.29 (as Safety Gtiides 26 and 29 on March 23 and June 7,1972, respectively). Also,

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the proposed IEEE Std 279, dated August 30, 1968, was not used in the design 1of the facility. Therefore, for this evaluation, systems which should be

" safety-grade" are the shutdown systems classified in Table 3.1 and those tabalated in the minimum list of safe shutdown systems that follows.

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General resign Criteria (GDC) I through 4 (2] require that systems, structures, and ccmponents important to safety (1) be constructed to quality standards, and (2) be protected from the effects of natural phenomena (earthquakes, etc.) and other conditions (fires, pipe breaks, etc.). GDC 5 requires that systems important to safety not be shared among other nuclear powcr units unless such sharing does not significantly impair the performance of system safety functions.

For Yankee Rowe systems and equipment, the various aspects of GDC 1 through 5, including the systems required for safe shutdown, will be evaluated elsewhere under several SEP tcpics.

In order to acccamplish a plant shutdown and cooldown following a loss of of fsite pcwer, certain " tasks" must be performed, such as core decay heat removal, steam generator II.akeup, and ccmponent cooling. The staff and Licensee develcped a " minimum list" of systems necessary to perform these tasks, considering a loss of of fsite ac power and the most limiting single failure. Although other systems may be used to perform shutdown and cooldewn O

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TER-C5257-310 functio ns, the following list is the minimum number of systems required to fulfill the BTP RSB 5-1 criteria:

1.

steam relieving paths involving =ain steam, auxiliary steam, and heating steam systems.

2.

auxiliary feedwater systen 3.

water sources (demineralized water storage tank, primary makeup tank, and safety injection tank) 4,. shutdown cooling system 5.

component cooling system 6.

service water system 7.

pressure control and relief system-

,8. chemical and volume control system 9.

control air system 10.

emergency pcwer systec 11.

instrumentation for shutdown and cooldown.*

"he staf f's evaluation of each of these systems, with respect to the BTP 5-1 functional requirements, is given in Section 3.2.

The power supplies and.

location of major safe shutdown components are also provided.

Tcr a list cf safe shutdewn instrumentation, see Section 3.3.

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liP on TABLE 3.1 CLASSIFICATION OF SilUTDOWN SYSTEMS - YANKEE ROWE 9;;

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e. 3 g3 Quality Group Selanic Plant Plant f

Cun ponent s/Subsys t eins R.G.

1.26 Design R.G.

1.29 Design Resnar ks in 3

Miln Steam Systein 5

Main steam headers ASMd III Note 1 Category I Note 1 frun steam generators Class 2 up to and including the EDFP, the 18-inch turbine throttle valves, the bypass valve, and y

connecting piping up to y

and including the first valve that is normally closed or capable of autc;aatic closure i

t Mnergency lloller Feed Planp ASME III Mfr. Std.

Category I Note 1 EDFP piping f run dis-ASME III ASA D31.1 Category I Note 1 charge of ptaup to main Class 3 feed lines including EllFP relief of Main feed piping from ASME III ASA D31.1 Category I Note 1 5

and including valves Class 3 b

3 MOV-1003 through 1006, U

CV-1000A, CV-1100A, U

b CV-1200A, and CV1300A, up to valves CV-1000, 1100, 1200, and 1300 9

r_.

~

TADI.E 3.1 (Con.t i riened )

!hd Quality Group Seismic gg Plant Plant 15' Canpo.ient s/ Subsystem s R.G.

1.26 Desijn R.G.

1.29 Design Remarks

+g Main feed piping from ASME III ASA B31.1 Category I Note 1 29-and including CV-1000, Class 2 p

1100, 1200, and 1300 y

up to the steam generators and connecting piping up to and including the first valve that is normally closed or capable of autunatic closure t1wP piping from suction ASMS III ASA B31.1 Category I Note 1 Refer to Technical T of pump to and including Class 3 Specification 3.7.1.3 5 the DWST and/or the PWST and connected piping up to and including the first valve that is either i

normally closed or capable of autanatic closure Shutdown Cooling System Pump ASME III Mfr. Std.

Category I Note 1 Class 3 ileat exchanger (shell side)

ASME III ASME VIII Category I Note 1 Heat exchanger also clase 3 (1956) constructed in accordance with'the 1956 edition of (tube side)

ASME III Standakds of the Tubular Class 2 Exchanger Mfr 's. Asso-clation

TABLE 3.1 (Continued)

.g p==

Qaality Group Selanic Plant Plant g

it 5' canponents/Subsy stems R.G.

1.26 Design R.G.

1.29 Design Remarks 2A'i SCS piping froin MOV-552 ASME III AS A B31.1 Category I Note 1 through the SCS pump and Class 2 (1955)

]yp heat exchanger to MOV-551 Sect. I and 6 j

! ?,

and connected piping up N

to the first normally closed valva or valves capable of autanatic closure Cunponent Cooling Mater

[

Planps (2)

ASME III ASME VIII Category I Note 1 tn Class 3 (1956) lleat exchangers (tube side)

ASME III ASME VIII Category I Note 1 (shell side)

Class 3 (1956)

CCW piping and connected ASME III AS A B31.1 Category I Note 1 Notes Piping which piping up to and including Class 3 (1955) Sect.

penetrates up to contalunent 1 and 6 the outermost con-the first valve that is tairanent isolation valve should be ASME either normally closed or capable of autanatic III, Class 2 closure g

o A

h, CCW surge tank ASME III ASME VIII Category I Note 1 Class 3 (1956) y U

CCW valves and fittings ASME III AS A B16.5 Category I Note 1 0

Class 3 (1957)

TABLE 3.1 (Continued)

Quality Group Selonic Plant Plant F

C< su ponent s/Subsy s t em s R.G.

1.26 IMs ign R.G.

1.29 Design Remarks

?$

h$

Service Water System e 5'

{$[

Puinpa (3)

ASME III Mfr. Std.

Category I Note 1 flass 2 0

I Q SWS piping and connected ASME III ASA B31.1 Category I Note 1 Notes Piping which Ej piping up to and including Class 3 penetrater. containeesit the first valve that lu up to the outermost contalanent isolation either normally closed or valve should be ASME III capable of autanatic Class 2 closure Pressure Control l

w

]

8, Solenoid-oper a ted ASME III B31.1 Category I Note 1 pressurizer relief Class 1 ASME Sec I m

valve and B16.5

(

i Pressurizer relief ASME III B16.5 Category I Note 1

)

valve Class 1 i

k Main pressurizer spray ASME III D16.5 Category I Note 1 flow isolation valve Class 1 Category I Note 1 i

Pressurizer heaters N/A N/A i

Chemical and Voltane Control FSAR Sections 203 and 204 Systuu Planps (3)

ASME III ASME III Category I Note 1 Note: The system boundary class 2 (1956) includes connectieg piping up to and including the low pressure surge tank ASME III Note 1 Category I Note 1 first valve that is either class 2 normally closed or capable of automatic closure 1

9 1

4 i

ti,

!=

Q TABI.E 3.1 (Continued) hh a3 Quality Group Seismic Plant Plant l3 sh C<anponen t s/ Subsystem s H.G.

1.26 Design R.G.

1.29 Design Remarks 0

t y

Piping and valves frun ASME III ASA 31.1-Categcery I Note 1 pump discharge to Class 2 (1955) Sect.

Cll-V-617 and CII-V-611 1 and 6 4

Piping f ran and inclu-ASME III Note 1 l

Category I Note 1 ding Cll-V-617 and Class 1 Cil-V-611 to the main coolant systan i

H

-4 I,etdown piping from the ASME III Note 1 Category I Note 1 main coolant system to and Class 1 including the orifice I

isolation valves i

Feed and bleed heat ASME III ASME VIII Category I Note 1 exchang ers Class 1 (1956) 1 I.etdown piping frun ASME III ASA 31.1 Category 1 Note 1 l

orifice isolation Class 2 (1955) Sect.

valves to pump 1 and 5 suctions via I.PST Piping from safety ASME III ASA 31.1 Category I Note 1 Note:

Ik, ration is injection tank to Class 2 (1955) Sect.

performed by CVCS pumps charging pumps via 1 and 6 using borated water fran MOV-540 up to and SI tank or BAMT including valve CS-V-630

TABLE 3.1 (Cent inued)

~~

Quality Group Seinnic Plant Plant Components /Subsystens R.G.

1.26 Design R.G.

1.29 Design Hemarks Piping from CS-V-630 ASME III ASA 31.1 Category I Note 1 BAMT was fabricated to

?g to tx)ric acid mix tank Class 3 (1955) Sec t.

ASME VIII 6

nE MOV-529 1 and 6

a. 3 i

p*y as CVCS valves and fit tings As atx)ve ASA B16.5 Category I Note 1 j

4 for piping (1953)

Sr neergency Power System g

a Diesel generators (3)

NA Category I Note 1 DC power system Category I Note 1 l

Distt it>ution lines, Category I Note 1 switchgear, control lx)ards, motor control l

7r centers I

Diesel generator ASME III Note 1 Category I Note 1 f uel oil systan Class 3 Control Air i

Air coupressors and Quality Note 1 Non-Seisn ic Note 1 Air sys; Ns required to associated equipnent Group D Categury perforia safety functions (e.g., acctanulator and piping to a safety-related valve) are seianic category I.

Service Air l

Air cumpressors Quality Note 1 Non-Se i smic Note 1 Group D Category l'

Note la Plant design information is not known.

I

TER-CS 257 -310 3.2 Functional Recuirement STEM 4 RELIEVING PATES Task: Removal, of core decay heat and main coolant system sensible, heat by venting steam frem the main steam system directly to at:casphere.

Discussion immediately af ter the loss of of fsite ac power, turbine trip, and reactor se' ram, the main steam safety valves automatically actuate to control steam system pressure and main coolant system temperature. However, the main steam safety. valves are not normally used at pressures Mlow their lif t pressure, although a lif ting lever is furnished oc each valve for manual operation. The cooldown of the Yankee Powe main coolant syste, following a loss of of fsite ac power would be accomplished using the atmospheric dump valve (ACV) and several other steam flow paths. The following paragraphs will briefly describe each vent path.

The air-controlled ADV* vents steam frem any of the four 14-inch (outside diameter (CD]) main steam lines between the vapor containment and the turbine building. The ADV vents steam frem a st.eam header pressurt:ed by manually operated 1-inch isciation valves from any or all of the four main steam lines. Tb? piping system is arranged such that the ALV can remove energy rrom any or all steam generators.

The Licensee calculated the capacity of the ADV based on 775 psig saturated steam and critical flow. The mass flow rate out the ADV 'is about 29,500 lbm/h or about 9100 Stu/sec (based on 1199 Stu/lbm hg and 88 9tu/lbm h **).

g

Air is supplied to the ADV diaphragm from either the instrument air system or from a newly installed (dedicated) N2 bottle.

- The staf f notes that most of the feedwater inside the steam generator is at T

and therefore an hg of about 500 Stu/lbm would have been more

rot, appecpriate initially. However, even use of this enthalpy does not acccunt for the diff erence between the calculated and measured energy removal rate of the ALv.

3 -19 E_nklin Research Center

(

T ER-CS 257-310 Actu'al measurements, however, indicated that the capacity cf the ADV is considerably less. The tests were conducted by calculating the cooldown rate with the ADV fully open, knowing the heatup rate with the ADV shut. The tests I

showed the ADV able to remove only about 3100 Stu/sec.

Based on actual measurements of these tests, the Licensee considered it necessary to provide another. flow path for energy removal. The flow path created allows steam to pass from the pressurized steam header, as with the ADV, through two manually operated valves (AS-V-720 and AS '?-721) and then to atmosphere through a 1-inch pipe.

AS-V-720 is normally open, while AS-V-721 is normally closed. The Licensee's calculated energy removal rate using tnis path is about 9661 Btu /sec. This value was calculated by assuming a steam pressure at 775 psig, critical flow from a 1-inch pipe, and a feedwater inlet of 120*F.

The plant operating procedure for a " Loss of A.C. Supply," OP 3251

' discussed in Section 2), directs the operator to start the emergency boiler feed pump (ESFP) and lit 19 steam to the large " hogger." The large and small l

hoggers at Yankee Rowe are single-stage vanturi-type air ejec, tors which draw from the condenaar and exhaust directly to atmosphere (unlike the main air ejectors which exhaust to the shell side of a conder.ser cooled by condensate).

The hoggers are normally used for removing large amounts of air and gasses from the condenser during startups.

(During startups, steam for the hoggers ccmes from the main steam lines.)

The hoggers can be used to remove energy from the steam generators by bleeding steam from the main stem system to the hoggers via the auxiliary -

steam system.

In this mode of operation, the suction valves to the main condenser are shut. Main steam is throttled at the nozzle inlets to maintain about 300 psig on the large hogger and 60 psig on the small hogger. Since there is no automatic pressure regulater, as main steam pressure drops during the reactor coolant system (RCS) cooldown, the throttle valve setting must be manually adjusted.

The EBFP is utilized to provide feedwater to the steam generators during the less of of fsite ac power and is described in the following section. The

&s 50 Frank!in Research Center 3-20 m notner -.=w.

TER-C5257-310 EBFP can be used to remove energy from the steam generators, since the EBFP uses steam from the main steam system via the auxiliary steam and heating steam systems. Steam pressure is automatically maintained at 100 psi by pressure control valve PCV-305.

The earliest time following a lo'ss of of fsite ac power and reactor shutdown when each component's energy release rate equals the core decay heat generation rate is provided below, i

Energy Bemoval Rate Comoonent (Btu /h)

Time thours) 6 ADV 13.5 x 10 s10 6

1-in vent 4 2.3.x 10

%.2 6

Large hogger 5.50 x 10

>36 l

Small hogger 1.12 x 10

>36 ESF?

2.27 x 10

>36 Staff scoping calculations assumed that steam generator pressure remained at 935 psig (lowest main steam safety valve setpoint) until the component energy removal rate equals the decay heat generaticn rats. The time l

l calc'; lated represents the approximate time when (1) plant cooldown commences if the component is used and (2) intermittent main steam safety salves lifting would stop.

Redundar.cv To establish the degree of redundancy provided by the various components l

discussed above, the staff and Licensee calculated the main coolant system cooldown times using various ccmbinations of the components. The staff's calculatiens are summari:ed below:

Comoenent(s)

Pasults ADV 494*F in 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months 1-in vent 370*F in 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months ADV + l-in vent 351*F in 50 hcurs (s 330*F in 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months )

ADV + l-in vent + EBFP 330*F in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months B-21 Ah Frantra Research Center A W af *he Frarwe., arisonme I

b

TER-C5257-310 Using these results to establish the redundancy, it is apparent that, even if all steam vent paths are considered, the main coolant system cannot be cooled down within a reasonable period of time (defined as 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months in Standard Review Plan 5.4.7). A single failure within the vent paths of the ADV or 1-inch vent would extend the time required.

The staff also performed scoping calculations to cetermine the dependence of RCS cocidown time on the initiation time. It was found that, if the cooldown were delayed 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the time to reach 330*F would be the same as if the cooldown began immediate!.y (as soon as possible af ter the scram). Since

'the core decay heat is less at 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and the energy removal rate is the same, the cooldown rate is higher initially, but then decreases as the energy removal rate (determined mainly by steam, pressure) decreases.

Although the assumptions and calculating methods varied between the

!ticensee and staff analysis, the Licensee's results support tne staff conclusion that the steam vent paths do not have sufficient capacity cr redundancy to satisfy the -functional requirements of BTP RSB 5-1.

In a March 26, 1981 letter (3), the Licensee proposed changss to provide,autcmatic quick closure of the fcur main steam line non-return valves. This modification necessitated the installation of a new steam supply line to the steam-driven emergency feedwater pump and installation of additional steam dump capacity.

During a March 27, 1981 discussion (4], the Licensee indicated that an additonal manually operated dump valve would be installed on each steam line upstream of the ncn-return valve. Each of these valves is to have tha ability to remove approximately 60,000 ihm/h. The Licensee stated that the new dump valves are intended to be the main method of decay heat removal following the loss of ef fsite ac power.

The staf f performe. scoping calculations to assess the plant cooldewn capability based en the preposed modification. A single failure was postulated to one manually cperated dump valve and a decay time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to ecmmencing cooldown. Based en these assumptions, a main coolant system e.emperature of 330'? was attained in appecximately 8.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . The results demonstrate that the proposed modifications as described would affctd

~h :bD. 'in Research Center B-22

.... ranx 4emaa o.r n am.ou.

m.

~.

i TER-CS257-310 suff~icient capacity and redundancy to satisfy the functional requirements of BTP RSB 5-1.

,e Based on the above discus:,ien, the staf f concludes that the current steam relieving paths do not conform to Criterion 4 and that preposed modifications as described would provide sufficient capacity and redundancy to satisfy the fonctional requirements of BTP RSB 5-1.

The need for design modifications to provide additional steam relieving capacity over and above the current capacity will be addressed in the SEP integrated assessment of the plant.

Location and Oceratien The staff evaluated the equipment discussed above with respect to its 1ccation and operability during a loss of offsite ac power. Table 3.2-1 shows the equipment's location, the points from which it may be operated, and its ipower supply. The design of the electrical instrumentation and controls for this and cther safe shutdown equipment will be evaluated in the electrical portion of the staff's review of Topic VII-3.

At:XILIARY FEEDWATER SYSTIM Task: Provide 5 team generator makeuo inventory whenever the main coolant sistem temperature is greater than 330*F and the feedwater system is inoperable.

Discussion

While the main coolant system temperature is above 330*F, the core decay heat is removed by bleeding steam fecm the steam generators using the various fiowpaths discussed in previcus sec.icns. The condensate and feed pumps are powered from the 2400-v bus which is normally supplied from offsite power.

Following a less of offsite ac power these pumps will not be available.

Pre-core XV co.nfiguration 4%

-.UJ5nklin Reses.% a Center 3-23

= o~,. = w w re. -.

i t.

lf*

b 7:

Y!

Ic s 5' Tablr. 3.2-1 UN l4

?

EQUIPMENT IOCATION CONTitOL POINTS ELECTRIC j

in I$

Large and mall Turbine building,.adja-Incal manual operation oaly h electrical power is needed.

O hoggers cent to condenser hotwell (Open/ shut steam inlet valves.

about 50 ft from feed-s water regulating valves, and one flight of stairs 2

below control room.

Atmosphere dump Outside, in the vicinity Control roon operation and No electrleal power is needed.

4 valve of the MSSVs, accessible Local operation using nitrogen by catwalk about 30 ft bottle pressure 8a the lower 4-above ground level.

level of the tu. 4ne building.

1-in vent pipe Valves to lineup to con-Inc<.; operation only using the No electrical power is needed.

trol this path are manual control valves, located in the heating boiler roan; the 1-in i

vent pipe goes to atmo-sphere just outside the boilder room.

HIFP See following section.

See following section.

3 N

A C

TER-CS 257-310 The auxiliary feedwater system is designed to supply water to the steam generators for main :oolant system decay heat !?moval when the normal feedwater system is not available. The auxiliary feedwater system is not normally used for other plant operations such as startup or shutdown. The auxiliary feedwater system is initiated by starting the emergency boiler feed pump (EBFP) locally and by opening four normallf closed manual valves in parallel discharge lines to each steam generator. The valve operation is also accomplished locally. Af ter s, tarting the pump and opening valves, the flow can be controlled from the control room.

The EBFP is a turbine-driven reciprocating pump that provides a minimum of 80 gpm directly into the four feedwater lines immediately upstream of the air-operated feedwater regulating valves (FRVs). Steam for the turbine is supplied from the auxiliary steam header, which is supplied from either main I

steam header via an automatic reducing valve or from auxiliary boilers.

Turbine exhaust is directly to atmosphere via the vent stack. The ESFP is i

lined up to receive water from the demineralized water storage tank and can also receive water directly from the primary water storage tank. The EBFP discharges to the main feed system via a 2-inch (CD) header. The header 1

divides into four 1.5-inch (CD) lines, each of which pressurizes one of the 1

four normal feed lines downstream of the motor-operated isolation valves (M3Vs). Each 1.5-inch (CD) line has a manual isolation valve that is opened to pressurize the four feed lines. Steam generator level control is performed l

using the individual FRVs* from the local statien after the four McVs are shut.

l l

Redundancy Since the auxiliary feedwater system consists of a single ESFP and piping train, it is susceptible to single failures. A backup method of supplying feedwater to the steam generators in the event of failure in the auxiliary feedwater system is the charging pumps with a total capacity of approximately

'The FRVs are ncrmally air operated but can be manually operated (during a less of instrurtnt air) using a handwheel.

4 NJ Franklin Research Center B-25 a o, a e n. r,.a m

~. -

O e

TER-CS 257-310 100'gpm.

Two of these pumps have variable-speed motors. The charging and volume control system (CVCS) is connected permanently by a spool piece to the feedwater system. The operation of ten valves (two drains and eight isolation valves including manual valves CH-V-69 2, CH-V-751, CH-V-64 2, CH-V-641, ' and CH-V-689) is required to initiate fl,ow from this source. The CVCS is also connected to the steam generator blowdown piping. Manual operation to open valves CH-V-741, VD-V-1093, VD-V-1094, VD-V-1095, and VD-V-1096 is required to establish a feed path to the steam gnerators. Both of these paths use

~

non-nuclear system (NNS) piping. The water supply to the charging pump is the 135,000-gallon primary water storage taaA or the 30,000-gallon demineralized water riarage tank.

The high pressure safety injection -(HPSI) and low pressure safety injection (LISI) pumps provide two additional methods of supplying feedw;ter t'o. the steam generators. The first path is from the safety injection discharge header through normally closed motor-operated valves SI-MOV-514 and

-515 and manual throttle valve SI-V-645 to the charging header. Therefore, safety injection water can be directed to t'_- charging header, and di.stributed to steam generators by the CVCS connectio.s te :he feedwater of blowdown piping. Discharge from the EPSI and LPS. punps can also be directed through manual valves SI-V-700, VD-V-1093, VD-V-10.4, VD-V-1095, and VD-V-1096 to establish a feed path to the steam generators through the blowdown system.

The flow available from the EPSI and LPSI sources is 200 gpm per train (three trains available).

Power for the charging pumps and motor-sperated valves is supplied.' rem separate nonsafety 480-V ac buses, which are capable of being fed by the i

emergency 480-V ac buses by remote manual cperation of circuit breakers. The EPSI and LPSI pumps are connected to the 480-V emergency buses. Following a loss of of fsite ac power and a single failure in the auxiliary feedwater system, the charging pumps would not be available unless operator action occurs to initiate manual operation af circuit breakers to supply emergency power. In a December 21, 1979 letter (5), the Licensee concluded that there is not suf ficient emergency diesel capacity to provide normal power to the charging pumps and simultaneously supply the existing emergency core cooling i

nkun Research Center B-26 A Omsen cd he F enen esonne j

I l'

TER-C5257-310 require-

.es. Since the charging pumps are not supplied from emergency buses, i

the safety injection system may be required to fulfill functiens normally i

assigned to the charging pumps (i.e., bor ation and primary plant makeup). In I

addition, the safety injection tank functions as a scuree of borated water similar to a refueling water storage tank in other Westinghouse' reactor t

facilitiaa. If the safety injectien tank is used as an alternate source of l

water, the dissolved boron will be concentrated in the steam generators by the release of steam. The volume of water in the safety injection tank cannot be co'nsidered as an alternate source of water for the steam generators except under very extreme conditions. Censideration of severe conditions warranting such use of the safety injection tank is outside the intent of the safe shutdown review.

Since the charging pumps cannot be considered available as a backup bethod to feed the steam generators following a loss of offsite ac power and the safety injection system does not have a suitable source of wate-for steam generator feed, the auxiliary feedwater system as currently designed does not satisfy the functional cr'iteria of BTP RSB 5-1.

In Reference 5, the Licensee described propose 3 auxiliary feedwater system design *htnies to provide redundancy to the system. As proposed, the revised sys' ists of two 100-percent, safety. class electric pumps. driven from redundant powsr sources. The preferred flow path is to the existing l

auxiliary feedwater header. An alternate flow path is proposed utilizing the containment penetrations provided by the steam generator blowdown pipes.

Check valves in the blowdown lines direct auxiliary feed to the feed nozzle and prevent flow from entering the steam generator at the blowdcwn connection. The new pumps, located in the primary auxiliary building, are to be capable of being started either locally or from the centrol room.

The Licensee has indicated that the existing steam-eperated ESFP would be retained but its intended emergency function would be modified to mitigation of station blackout only. In addition, before the new auxiliary feedwater pumps are made operable, it is necessary to upgrade the onsite electrical distribution system to acecmmodate the additional electrical Icads. The l

l 3-27 E,.nklin Research Center 4 o

a. v.aw.a w

TER-CS257-310 existing emergency diesel generator capacity is to be augmented to provide complete operability of the new auxiliary feedwater pumps.on loss of station power.

sed on the above discut.icn, the staf f concludes that the current auxiliary feedwater system does not meet the functional requirements cf BTP RSB 5-1 but that proposed nodifications as described would satisfy the functional requirements of BTP RSB 5-1.

The need for these design modifica-tiens to provide a more reliable auxiliary f.adwater' systara #11 be addressed in the SEP integrated assessment of the plant.

Iceation and operation The staff evaluated the equipment *tscussed above with respect to its location and operability during a loss of offsite ac powers Table 3.2-2 shows

the equipment location, tha points from which it may be operated, and its power supply.

WATER SCUF"" (CEMINERALI::ID WATER STORAGE TANK, ORIMART MAKEUP TANK, AND SAFETY INJECTICN TANK)

Tssk: Provide a sou ce of auxiliary feedwater, primary makeup, and bcrated water.

Discussion The E3FP takes a suction from the demineralized water storage tank (OWST) via a 10-inch (CD) line which also serves as the betwell aakeup and rejection lir.e. This line leaves the bottem of the CWST and frem there branches inte the following:

1.

a 10-in betwell rejection line (i.e., flew from hetwell using condensate pumps and a level control valve) 2.

a 10-in hotvell makeup line 3.

a 3-in E3FP suction line 4.

a 4-in LPST makeup and charging pump suction line 5.

a 4-in auxiliary boiler makeup line.

O

..i. Frankin Research cente, a-28 4 w n.rr aw u.

i

~

4 b.

~

.;* g Table 3.2-2 l

e,. a CONTitOb POINTS ELtrutIC i

(f.

IlX' ATION EQUIPm.HT N

anergency boiler IM corner of heating local operation only. Once at to electrical power is needed.

o boiler room floor, which proper rpm, governor maintains feed pump is a partitioned part of speed.

the turbine building.

~

No electr ical power is needed.

Charginu to feed Spool pieces and piping local manual only.

system spool flanges (and bolts) are in charging ptsap cubicle.

piece Valves connecting to feed systesa must be opened in as O

lower level of turbine building, in vicinity of EBFPs (6 ft north of EDFP motors).

Charging ptmaps Ptsaps are located in Ptsaps operated froia control CPS 1 - NCC 4, Dus 1 (480) and valves separate cubicles in roon or locally at their CPS 2 - MCC 2, Buu 1 (480)

PAu. valves are under controllers (oper door and use CPil - HCC 4, Bus 2 (4GO) pan floor, with reach jumpers). Valves, are local manual only.

rods.

See unergency power g

E tr',

See usergency power See uncrgency power systen discussion systen discussion system discussion g

below.

A below.

below.

U S

J.

G

TER-C5257-310 Condensate and demineralized water are stored in the DWST and the primary water storage tank (PWST). The CWST is an aluminum 30,000-gallon tank chat is normally filled from the water treatment plant. The DWST is sized to handle all expected transients in the condenaate/feedwater systen. This is accomplished by providing makeup to and accepting rejected water ' rem the

~

condenser hotwell.

The EBFP can also take a succion from the PWST via a 4-inch (CD) line which airo serves ss an alternate supply of water to the charging pumps. The FWST provides deminerali:ed water for the primary plant as well as for varidus

' demands in the primary auxiliary building, the radwaste building, and the spent fuel stcrage area.

It is the supply for the low pressure surge tank makeup pumps and, as such, serves the above areas. The PWST is constructed of aluminum and has a capacity of 135,000 gallons. An inner floatino roof pejvents aeration of the tank contents. The tank receives makeup water directly from the water treatment plant.

Technical Specification 3.7.1.3 requires there to be a minimum combined volume of 85,000 gallons available from the PWST and the DWST.

The service water systam discussed later), which receives fresh water from Sherman Fond, supplies the water treating (WT) plant for PWST and DWST makeup. The WT plant

is sized to provide 40 7pm of deminertlized water on a continuous basis and 80 gym maximum, based en the average chemical analysis of Sherman Pond water obtained over a 1-year period.

The safety inj ection tank (SIT) is sized to provide a source of borated water to the saf ety injection pumps following a loss of coolant accident.

Another functicn cf the SIT is to provide a source of water for ficodir, the shield tank cavity curing refueling cperations.

The SIT also provides a source of borated water for the reactivity control system. Technical Specification 3.1.2.11 (Limiting Condition for Operation of Borated Water Sources) requires that the SIT be operable with:

'The WT plant is not included in the list of " minimum systems" but would j

probably be avai.'.able since it is essentially a passive system which is pressurized using the service water system.

B-30 l

_nkun Research _ Center

TER-C5257-310 e

a minimum contained borated water volume 117,000 gallons of water, equivalent to a tank livel of > 25.5 feet a minimum boron concentration of 2200 ppm e

o a minimum solution temperature of 40'F.

Fedundancy The staf f calculated the maximum length of time the plant can stay at hot shutdown following the loss of of fsite ac power, using the initial steam generator water inventory and a maximum CWST level or 30,000 gallens. The calculations show that approximstely 20.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months water supply are svailable for main coolant system temperature control; af ter that, the EBFP suction must be shif ted to the PWST. The staff also calculated that the total water inventory required by Technical Specifications (85,000 gallons)' is enough to keep the plant at hot shutdown for about 73 hours8.449074e-4 days 0.0203 hours 1.207011e-4 weeks 2.77765e-5 months . In Appendix B Part 2, Safe Shutdown Water Eequirements, the staff determined the time required to comp 1ste a shutdown to the point of shutdown cooling system operation. Assuming (1) no credit for the initial steam generator inventory. (2) no condensate in the hotwell, (3) no single failure, and (4) use of the atmospheric dump valve, the 1-inch vent, the large and small hoggers, and the EBFP (all steam ver.t paths),

the staff determined chat a water inventory of 72,000 Sailons is sufficient to conduct the ecoldown in 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Further calculations show that, if the plant stayed at hot shutdown for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and then plant cooldown was initiated, the cooldown rates would be higher, but the time to cool the main coolant aystem to 330*F would remain the same.

Since the condensate pump noter is much smaller than the boiler feed pump motcr (250 hp versus 700 hp), a single condensate pump can be started to pump the centents of the condenser hotwell back to the CWST for ESFP usage. The EBFP would not have to be stopped curing t:ds operation since its suction would just be augmented by the condensate parp (the condensate pump rejection line would pressurize the E3k'P suction and fill tho DWS;). The hotwell has a capacity of 15,000 gallons and a normel operating level of about 10,000 gsilons. However, the hotwell contents following a loss of offsite ac power and uabsequent feed and condensate pump trips cent.ot be predetermined since D

3-31

'.M Franklin Research Cen'er A Cnwe,an of S. Fenneen wetsees I

TER-C5257-310 event and component coast dcwn time cannot be accurately predicted.

a.terefore, no credit can be given for this inventory; however, it is likely that there would be a significant quantity of condensate available and usable.

The basis of the technical specifications for reactivit; control systems j

states that tae maximum beration capability requirement occurs at the end of core lif e from full power equilibrium xenon conditions and requires 9,192 gallons of 2200 ppm borated water from tho safety iMeetira tank. Since 11,7,000 gallons of 2200 ppm borated water is available in the SIT, the staff concludes that sufficient berated water capacity it provided to satisfy BRP RSB 5-1 functional requirements.

The amount of main coolant system makeup during the cocidown 'and filling of the pressurizer) frem 539'F to 330*F 'was calculated by the r'.aff to be about 6,000 gallons. Since the cooldown to shutdc.rn cooling system initiation used 72,000 gallons and since 85,000 gallons is available per technical specification, the staff concludes, that sufficient primary makeup water is available to satisfy B7P RSB 5-1 functional requirements.

Iccation and Ceeration The staff evaluated the equipment discussed above with respect to its location and operabili$y during a less of offaite ac power. Table 3.2-3 shcws

~

the equipment's locatiun, the points from which it may be operated, and its power sup;,y.

ShTTDCWN CCCLING SYSOD4 Taskt Removal of core decay heat and main coolant system sensible heat to cool the system from 330*F to 140*F.*

Discussion The rhutdown cooling system is placed in service af ter the main coolant temperature has been reduced to approximately 330*F and the pressure to less D

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TER-CS 257-310 than 300 psig. The shutdown cooling system then reduces the main coolant tetrerature to 140'? or less and operates continuously to maintain this temperature as long as is required by maintenance or refueling operations.

The shutdown cooling systtm consists of a heat exchanger, cir.culating pump, piping, valves, and instruments arranged in a low pressure auxiliary loop parallel with the main coolant loops. The r.hutdown cooling pump takes suction frem the hot leg of the main coolant piping on the reactor side of the loop stop valves and recirculates main coolant water through the tube side of the shutdown cooler and back into the cold leg of the main coolant piping,

'which is also on the reactor side of the loop stop valves. The main coolant is contained in a closed system, and reactor decay heat load is transferred through the shutdown cooler to the component cooling system wnich in turn is cooled by river water. This arrangement of providing the intermediate cooling medium of the ecmponent cooling system was selected in order to assure that any possible leakage of radioactive main coolant would not enter the river water.

The shutdown cooling system is cesigned to remove the re. actor decay heat about 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after shutdown following 10,000 full power hours of operation.

6 According to the License, s estimates, about 16.2 x 10 Stu/h are generated by the reacter and transferred to the main coolant system.*

Redundancy Although the shutdown cooling ' system censists of a single cooler and cooling pump, a complete tackup of this system is provided by the low pressure surge tank pump and cooler. The coolers and pumps are identical. ine low pressure surge tank cooler and pump are connected in pa allel with the shutdown cooler and pump. By employing double valving in the inlet and outlet lines of the main coolant piping, any ccmbination of pump or cooler can be used to maintain decay heat removal.

iormally the shutdewn cooler and

'Draf t ANS 5.1 decay heat curve predicts a P/Po 2 0.008 at t = 18.0 x 103 6 Stu/h.

sec, or abeut 16.34 x 10 4&

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TER-CS257-310 pump are aligned to the main coelant system and the icw pressure surge tank cooler and pump are aligned to cool the low pressure surge tank. By manaal valve operatica, a failed component in the shutdown cooling cy; tem (SCS) can be repl' aced by a similar component in the low pressure surge tank cooling subsystems therefore, the SCS has redundancy of components. Because of the sharing of common as tion and discharge piping, the SCS is susceptiole to passive piping f ailures.

Based on the above discussion, the staf f concludes that the SCS satisfies th6 functional requirements of BTP RSB 5-1, except that the SCS pump and the LPST cooling purp are not powered frcm ilesel-r,upplied electrical buses. The staff will evaluate the significance of this in tae SEP integrated assessment of Tankee Rowe.

N ation and Ooeration The staf f evaluated the equipment discussed above with respect to its location and operability during a loss of offsite ac power. Tsble 3.2-4 shows the equipment's location, the points from which it may be operated, and its power supply.

CCMPCHENT COCLING WATER STSTDI Task: Provide cool'ing water to the SCS and/or LPST coolers and to other essential equipment.

Discussion The component coc'.ing system is necessary to remove reactor decay heat i

from the shutdown cooling system heat exchanger (or the low pressure surge tank cooler) and to provide cooling to equipment necessary for plant cooldown.

The component cooling system consists of two coolers, two circulating pumps, a surge tank, a chemical addition tank and associated piping, system and instrumentation piping, valves, fittings, and instruments. This equipment is connected to two main piping headers. One supplies vapor container "D

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S SCS and LPST Cubicles in tne PAB, The SCS and LPST cooling txamps SCS sump Bus 5-2 (480 V) cooling pumps lower level, are operated from the control LPST cooling pump Bus 6-3 and coolers roon and can be operated (480 V) locally during an energency by jumping the cell switch and using the test control switc;. at the switchgear us i

cubicle. The coolers re-quire local, manual valve operation such that operation t roen the control roaa is not possible.

SCS valves See Sections 4.1 and See Sections 4.1 and 4.2.

See Sections 4.1 and 4.2.

4.2.

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i TER-C5257-310 4

components, the other supplies equipment outside the containment. Independent lines, provided with isolation valves located outside the vapor container, are connected from the vapor container supply header to the various components inside the vapor container.

i A surge tank (4,000 gallons) is used in the component cooling system to provide makeup water for the system, to accommodate the expansion and contraction of the water in the system as temperature changes, and to act as a receiver for the safety valves in the camponent cooling lines. The water level in the tank is maintained at approximately 2,500 gallons. The surge tank is equipped with a vent to the primary vent stack' and a safety valve i

which discharces into the vapor container dr in tank.

Level controls and alarms are provided on the surge tank. A low pressure a,larm and pressure indicators are provided in the outlet of the component

cooling pumps. A ' pressure switch starts the standby pump on low pressure.

The common cooler inlet and outlet pipes are provided with local temperature indicstors, while the outlet pipe is also provided with a remote temperature indicator, a high temperature alarm, a flow metei, and a flow meter alarm. Controls for the component cooling pumps are located on the nuclea'r auxiliary panel in the main centrol room.

Two motor-driven centrifugal circulating pumps are provided. The capacity of each pump is approximately 2,000 gpm, with a total dynamic head of 190 feet of water and a design discharge shutoff pressure of 110 psig. The switches may be set in "Close," " Auto," or " Trip" position with provision for

" Pull-out" in the trip position.

l The two component coolers are of the shell and tube design and are provided to transf er heat from the component cooling water (CCW) to the l

service cooling water. The tubes are made of admiralty metal.

During a cooldown of the main coolant system following a loss of etdsite ac power, the SCS is used to circulate the hot main coolant through the SCS cooler (tube _ side). The shell side of the cooler is furnished with CCW, and b Franklin Research Center B-37 A Dmmen of The Frenman Inesome

TER-C5257-310 the cooldown is controlled by an air-cperated tenperature control valve *

(T/-200 ) on the CCW discharge of the I2S7 and SCS coolers (ccmmon line).

T/-200 controls either the LPST or SCS return (to the RCS) temperatute at 140*F by throttling the ecmpenent cooling flow from the coolers. khen the SCS cooler is first placed on line, the heat load is greatest and the SCS return temperature is highest, so the maxirrum CCW flow to the SCS cooler is allowed.

hten the SCS to RCJ temperature drops, then the CCW ficw is reduced to decrease the heat removal.

Redundancv Each cooler is designed for the full cooling capacity reached during 0

normal plant operation (8.5 x 10 Btu /h) and either cooler serves as a spare for the other; they can be cperated in parallel, if required. Also, each cooler can remove the maximum decay heat removed by the SCS cooler (16.0 x 0

10 Btu /h) with the er.me amount cf CCW flow.

Normal Ioad Full Icad Heat removal (x 10 Stu/h')

8.5

,16.0 CCW inlet temperature 100.5*F 96.0*F CCW outlet tarperature 92.0*F 80.0*F Service water inlet temperature 81.0*F 60.0*F Service water outlet temperature 87.8'r 72.8'F CCW flow 2000 gpm 2000 gym SW flow 2500 gpm 2500 gym Each pump can be operated singly er in parallel and is provided with a redundant independent power supply.

In additien, there are installed hose connections at the CCW pump discharge and suction to allcw service water er fire water to provide

r/-200 is ncemally air operated but fails open (during a loss of instrument air). Control of the CCW flow can thereafter be acecmplished using the manual isolation valve.

A dLb Franklin Research Center S-38 4 w.amr e.

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TER-C5257-310 component cooling should both CCW pumps be inoperable or if a rupture in the system has occurred. The peccedure (CP3115 - Loss of Component Cooling) directs the operator to attempt first to hook up to the portable fire hose from the fire system to the CCW system, then, if unable to use the fire system, to use the service water system. Thus, there are redundant and diverse means to provide component cooling.

Based on the above discussion, the staff concludes that the CCW system satisfies the functional requirements of BTP RSB 5-1, except that the electrical components are not powered from diesel supplied electrical buses.

The staf f will evaluate the significance of this in the SEP integrated assessment of Yankee Rowe.

Location and Ooeration I

The staf f evaluated the equipment discussed above with respect to its location and operability during a loss of offsite ac power. Table 3.2-5 shows the equipment's location, 'the points from which it may be-operated, and Ats power supply.

SERVICE WATER SYSTDi Task: Provide cooling water to the component cooling water coolers and the SCS pump and/or LPST' cooling pump coolers.

Discussion The service water system consists of three 2,500 gpm vertical deep well l

type pumps which obtain their suction from a common intake well in the circulating water pump house. The pumps discharge to a ecmmon 12-inch header which branches into two 12-inch supply headers. The supply headers run parallel to the southern wall of the turbine room basement. The two 12-inch supply headers furnish the various components with service water via separate aps from one or both of these two main supply headers. The headers can be 1

(manually) cross-connected so that any ccabination of pumps supplies the necessary loads.

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k EOllIPMENT IDCATION CONTitOL POINTS EL1Xllt IC 0

s CCW pumps (2)

Floor level of the Pall, Operable fran the control CCW pump 81 - Bus 83 (2400 V) gQ SW end of building, roon. Can be operated locally CCW pump 82 - Dus 82 (2400 V) under CCW surge tank, in an energency by jumping the cell switch and using the test control switch at the 2400 V breaker.

CCW coolers (2)

Side-by-side, upper level local-snanual operation of No electrical power is needed.

of Pall, SW end of build-valves.

w 1

Ing, adj acent to CCW surge tank.

CCW houe fittings Incated at. various places Incal-inanual operation only.

No electrical power is needed.

and portable hoses in the SW end of the Pall, I

all within about 50 ft.

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TOR-C5257-310 The greatest haat load on the system occurs when the SCS is first placed in cperation. A total of 2500 gpm of 60*F cooling water is required at that t ;m e. This same flow is required at any time when the main coolant systes v4ter chemistry requires operating the purification system at its maximum capacity of 100 gpm. There is adequate capacity in the service water pumps to meet these special cperating conditions.

Redundancy Nocnally, two pumps will be in operation, with one pump on standby.

If

'the pressure in the discharge header falls below a preset value, the standby pump will start and shnultaneously an alarm will be given at the main control board. The pressure switch for initiating this standby cperation is located in the turbine room and is set at approximately : psig.

I-The 2400-V power supplies to service water pumps fl, 2, and 3 See Bus 33,

~

Bus fi, and Bus 62, respectively. These buses can be separated so a fault in one would not disable any~ more than one service water pump.

Should all pumps fail due to electrical problems, localired damage in the screen house, or loss of suction from Sherman Pond or if a break aff ecting certain portions of the service water header should occur, selected service water loads can be provided with cooling water f cm the fire system. The fire system could be supplied by either the installed fire pumps or from portable fire pumps connected in series taking a suction frau the river or from Sheenan Pond.

Also, the potable water system can supply selected service water loads with cooling water. The plant procedure (CP-3009, Loss of Service Water) describes which components may receive fire water or potable water and the locations of the neces sary connections.

Based on the above discussion, the staff concludes that the service water system satisfies the functional require: tents of BTP RSB 5-1, except that the electrical components are not powered from diesel-supplied electrical buses.

The staff will evaluate the significance of this in the SEP integrated assessnent of Yankee Rowe.

B-41

_nklin Rese_ arch Center

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TER-C5257-310 Locatiu and Coeration The staf f evaluated the equipment discussed above with respect to its location and operability during a loss of of f:.ite power.

reble 3.1-6 shows the equipment's locaticn, the points from which it may be operated, and itr power supply.

1 PRESSURE CCNTROL AND RELIE SYSTDi Task: To maintain a system overpressure during het standby and/or natural circulation cooling aad to deprescurize the main coolant system tc

' permit the initiation of the shutdown cocling systen and to cool down the pressurizer.

Discussion I

f!.

The pressure control and relief system primarily functions to maintain the required 'nain coolant pressure at the reactor outlet during steady-state operation, to lhait to an allowable range the pressure changes caused by main coolant thermal expansion and contraction during nc: mal load transients, and i

to prevent the pressure in the main coolant system from exceeding the design pressure. The pressure control and relief system con 21ses of a pressurizer vessel containing a two-phase mixture of steam and water, hmmersion heaters, I

saf ety and relief valves, spray system, interconnection piping, valves, and 1

instrumentation.

I Depressurization of the main coolant system in pressurized water.eactors

(

is generally achieved by the pressurizer in con $2nction with one or more of the ic11owing:

(1) the main pressurizer spray, (2) the auxiliary pressurizer spray, or (3) the pressurizer relief valve. The pressurizer spray nozzle is located in the manway at the top of the prescurizer. The spray pipe is l

connected to the main coolant system inlet pipe en the reactor side of locp number 2 isolation valve. This connection is in the form of a scocp inside the coolant piping, so that the velocity head plus the static pressure dif ference between this connection and the surge pipe connection provide the maximum possible driving force tor spray flow. A motor-cperated valve on the spray can be cperated by a switch en the main control board.

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EOllIPMENT IDCATION OPEH7 TION POWER SUPPIN h

SWPs (3)

Circulating water ptamp Operable frosa the control SWP ll - Bus 83 (2400v) house roans can be operated SWP $2 - Bus 81 (2400v) locally in an energency by SW? 83 - Bus 52 (2400v) jumping the test control switch at the 2400v breaker 44 4

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Ilose connections Inlet and outlet to the Local-1aanual Ib electrical power is needed to SW systesa CCW coolers

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TER-C5257-310 Pc' lowing a loss of offsite ac power, the main coolant pumps are not available to sustain primary coolant flows therefore, normal pressurizer spray flow is not available for main coolant system depressuri=ation. As a consequerce, depressurization of the main ecolant system must be achieved with eithat the auxiliary pressurizer spray or the pressurizer solenoid-operated relief valve, PR-SOV-90.

The solenoid-operated relief valve can be operated manually by a switch in the control rocm. A motor-operated velve placed on the solenoid va2 ve inlet piping is provided for isolation of the solenoid valve. The steam and/or water discharged from the solenoid-operated relief

. valve passes through closed pipias to a blowdown eductor located in the water volume of the low pressure surge tank, which is part of the chenical and volume control systen.

The auxiliary spray line is located in the chemical and volume control system.

It is connected to the feed line devastreara of the f ted an1 bleed f,

heat exchangers. This, arrangement permits charging of water by the hich pressure charging pcmps into the tcp of the pressuri=er.

Follcwing a loss of of fsite ac power, main coolant system pressure control is necessary to maintain an adequate subcoeling ma: gin and assure no disruption of the natural circulation flow. Once natural circulation is achisved, system pressure control would be accomplished by maintaining a system overpressure with the pressurizer through use of the chemical and f

volume control system or pressurizer heaters. There are 48 pressurizer heaters combined into 24 groups. The 24 groups are combined into eight i

3-phase groups of 17.5 kW capacity.

Redund ancy In an April 9,1980 letter (6], the Licensee indicated that cperating expersence at Yankee Rowe has demonstrated that the operation of one group of pressurizer heaters (37.5 kW) is required to meet the heat loss from the pressurizer with nocmal spray flow through the pressurizer at hot standby conditions. The :bility to maintain natural circulatica under emergency conditions would require less capacity than for nonnal cperations. The Licensee. also indicated that Weetinghouse had performed a study to determine 1

l M.fE5nklin Research Center B-44 A m au m vmmene L_

TER-C5257-310 minimum heater re'"airements withovt offsite power and that extrapolation of the res>Tsts of this study to the Yankee Rowe pressuri:er confirmed the required heater capacity.

The Westinghouse study also determined that the capability to supply emergency power to the heaters within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months would prevent los's c2 subcooling in the primary system following a loss of of fsite pcwer. The Yanke. lawe facility has four groups of pressurizer heaters connected to 480-V Bus 6-3 and four groups connected to 480-V Bus 5-2.

Each of these buses is connected te an energency bus via two circui; breakers in series, and these tie breakers are operated from the main control room. To supply the required heater capacity from the emergency bus, sus 6-3 and Bus 5-2 are cleared and the buses are re-energized by closing the tie breakers to the emergency buses. The Licensee indicates that the tkne required to accomplish this, c.tving due Eonsideration to all requirements of plant operating precedures, is 15 minutes from the occurrence of the loss of of fsite power.

In a March 19,1981 ietter [7], the Licensee described design features of a preposed alternative safety shutdown system (ASSS). The Licensee indicated that the ASSS is designed in acccedsnce with requirements oi Appendix R to 10CTR50 as further clarified in the NRC Generic Letter 81-12, dated February 20, 1981. The proposed design har one group cf pressuri;er heaters being powered by 93 diesel generator such that main coolant system pressure can be maintained. One croup of heaters is capable of maintaining a hot shutdown ecndition.

As described, the pressure control and relief system has two methods of depressuri:ation. A single f ailure of the solenoid-operateJ relief valve or its blocking valve would nct preclude the capability to depressurize, provided auxiliary p essurizer spray ficw is available from the chemical and volume ucntrol system. The 3vailability of this flow path is assessed in the discussicn cf the chemical and voli:me cont-ci system.

Based on the above discussion, the staf f def ers evaluation of the adequacy of the pressure control and relief systen to satisfy BTP RSS 5-1 pending resolutien =f current staf f reviews of appli:able TM -2 action itecs k

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ten-C3257-310 and fire protection requironents. The staf f will deter:nine the effect that the conpletien of these reviews will have on the safe shutdown topic during the integrated assessment.

tocation and Oceration The staf f evaluated the equipment discussed above with respect to its location and cperability during a loss of offsite ac power. Table 3.2-7 shows the.quipment's location, the points from which it may be operated, and its po'wer supply.

CHD(ICAL MD VOLT 2iE CONTROL SYSTD4 Task Provide main coolant system makeup (due to the contraction of the coo? ant duririg the the cooldown) to provide a flow path for borating the main coolant system to the necessary shutdown margin, and to provide a means for depressuriration of the main coolant system.

Disn m.cn The chemical and volume control system censists of three positive displacenent charging pu:rps, feed and bleed heat exchangers, pressure reducing orifices, LPST, LPST pump, LPST cooler, LPST makeup pumps, and associated piping, valves, fittings and instruments.

i During nctmal operatien, bleed flow passes from No.1 locp T line, through the tube side of the feed and bleed heat axchangers, through the vc.r i-orif ice, and finally into the LPST through an eductor. Charging flow passes frro the purification pump discharge through the charging pumps, through t.he shell side of the feed and bleed heat est hangers, and into No. 4 loop T line.

In addition, charging flow can be lined up to the ir.dividual h

1 cops via te safety injection systsi.

Each charging punp is a positive displacement reciprocating pump rated at 33 gym, 2500 psig and driven by a 5.0-hp mocor.

No. 1 and 3 pumps have variable speed drives.

No. 2 punp is directly coupled to its meter and its O

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EQUIPMENT

_ILC ATION OPERATION POWER SUPPLY 3

PR-SOV-90 Vapor C.mtainer Control Hoon Battery 31 Solenoid Operated Pressurizer Relief f

Valve PR "OV-512 Vapor Container Control Roon 460V Eher9ency T

Pressurizer Relief MCC 1 O

Blocking Valve Pressurizer vapor Container Control Rocna Bus 5-2 (4 groups)

Heatura Bus 6-3 (4 groups)

PR-MOV-191 Vapor Cwtainer Control Hoon MOCl Bus 2 Main Presturizer s

Spray Flow Iso-lation Valve I

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TER-C5257-310 constant speed.

No. 2 pump could put out a variable flow by throttling CH-V-690 between the discharge and suction of the pump.

1 Charging pump suction can be from the following sources:

1.

LPST (gravity flow) 2.

Purification system (IX-gravity) 3.

Purification systen (pumps) 4, Borie acid mix tarrn (gravity) 5.

Saferty injection tank (gravity) 6.

WT systen (via LPST makeup pumps) 7.

PWST (gravity or LPST makeup pump) 8.

DWST (gravity).

Boration of the RCS is accasplist.cd by injecting borated water from I

either the boric acid mix tank (1500 gal at 12.0 to 12.5% by weight-min) or the saf ety injection tank (117.000 gal at 2200 ppm-min). The SI tank is nocnally used since it provides finer reactivity control because of its lower 4

bo'ron concentration; however, the boric acid mix tank is also available.

j A means for depressurization of the main coolant system is provided by the auxiliary spray line.

It is connected to the feed line downstream of the l

feed and bleed heat exchangers.

Tc initiate auxiliary spray' flow, meter-operated valves CH-MOV-524 and PR-MOV-191 Are closed and manual valve CH-V-6'13 is opened and throtti.d. CH-MOV-524 and PR-MOV-191 are operated remote manually frem the control roor.; however, access to the vapor container l

is required to locally cperate CH-V-613.

Fedundancy, To ensure that the pressurizer level can be centrolled during the most rapid cooldown (i.e., ensure sufficient charging pump discharge) the staff used the calculations of the main coolant ristLa cooldewn with the 1-inch vent af ter a wait time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This cooldown rate was initially (i.e., at

?

= $40'F) slightly greater than 50*F/h. The staf f calculated that the 3

liquid contraction rate due to the cooldown at about 50*F/h is less than the l

input rate available from each charging pump. Therefore, the pressurizer level can be raised by only one charging pump during this cooldown, and the renaining pumps provide further redundancy.

B-48

.... FranMn Research Center A Ommen of "M Fwmbn inesame

l 1

TER-CS257-310 i

Boration of the main coolant system cannot be accomplished at Yankee Bowe without-the charging pumps unless primary system pressure is reduced to allow use of the low pressure and high pressure safety injection pumps. The shutoff head of the high pressure safety injection pumps is 1850 feet, while that of the low pressure pumps is 1520 feet. Assuming that the low pressure safety l

injection pump is used as a booster pump for tne high pressure safety injection pump, the shutoff head of the safety injection system corresponds to approximately 1470 psi.

In order for the safety injection system to be a viable path for begated water addition, depressurization of the main coolant system would be required. As discussed in the previous section, two means are available to depressurize.

If a single f ailure is postulated in the pressure control and relief system (i.e., solenoid-operated relief valve does not open on demandi. the euxiliary spray flow from the chcaical and volume control system is required.

In order te ensure the availability of this depressuri-zation method, the charging pumps must be available.

The charging flow path is susceptible to single failure (e.g., operator error to close normally-cpen CH-McV-523 or CH-MOV-524), as is the letiown or bl.nad line (e.g., operator error to close normally-open CH-MOV-525 or i

LC7-222).

In addition, the charging pumps are not powered frat the 480-V f

emergency buses. The Yankee Ecwe facility has three charging immps powered i

f rom three diff erent 480,-V buses. Each of these buses is conne:ted to an f

energency bus via two circuit breakers in series, and these tie breakers are operated from the main control room. To supply the charging pumps from an energency bus, the non-emergency buses are cleared and then re-energired by closing the tie braakers to the emergency buses. In Reference 7, the Licensee proposed that a charging pump and associated motor-operated valves be powered by' #3 diesel generator tnrough a new 430-V ASSS motor control center.

Based en the above discussion, the staf f concludes that the chemical volume and control system does not meet the functional requirements of BTP RSB 5-1 in that the charging and lerdown paths are susceptible to single f ailure and that the charging pumps and other electrical components are not powered fecm diesel-supplied electrical buses. The staf f will evaluate the significance of these deviations during the SEP integrated assessment of Yankte Bowe.

B-G

_nklin Resead Center

,,m.

TER-C5257-310 Location and coeratien The staf f evaluated the equipment discussed above with respect to its location and operability during a loss of offsite power. Table 3.2-8 shows the ecuipment's location, the points from which it may be operated, and its power supply.

CCNTRCL AIR SYSTDi Tank: Provide compressed air for instrumentation and the control of' air-cperated valves in other safe shutdown systems.

Discussion Two 120 scfm, 600 rpm control air empressors with V-belt drive and 25 hp me". ors provide air at 100 psig to the instrument and control air system. Each

!empressor can operate in one of two modes. A local ecntrol switch with "Off-Hand-Auto" positions actuates the starter for each compressor. In " Hand" position, the empressor runs centinuously with the cmpressor loading and unloading autmatically to maintain rmeiver pressure. In " Auto" position, the ecmpressor motor is started and stopped autmatically to maintain receiver pressure. Each compressor is sired to provide 100% of the station's compressed ait requirements. The vertical, single stage, double acting, reciprocating, water-cooled compressors are of the non-lubricated carbon or teflon ring type and are installed with af tercoolers, air receivers, and intake f11ters.

)

The discharge from each control air receiver supplies one header of a double header piping system that runs throughout the station. The two control air neaders are cross-connected at the receivers in the turbine area and in the primary auxiliary building. Air fem each header is supplied through reducing valves, as required, to each instrument or control air supply manifold in the turbine area and primary auxiliary building. The reduced air station within the main control board has a low pressure alarm at 25 psig.

The control air header low pressure alarm is set at 75 psig. A solenoid-4 cperated bypass valve tha; opens at 65 psig connects receiver and header directly.

O OUbnklin Research Center 3~N

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TABI.E 3.2-8 nv l$

E001PMENT IDCATION OPEHATION POWER SUPPI.Y

s

.k Charging Punits (See EDFP discussioni)

(See IDFP discussion)

(See IBFP discussion) 83)

Boric Acid Mix typer level of PAB in Ideal manual o[eration only Mechanical agitator is [owered Tank general vicinity of the (filling, etc.)

from MCC4 Bus 1, trace com[onent cooling water heaters f rom HCC4 Bus 2, and surge tank redundant trace' heaters from energency tus 1.

Sa fety injection outside of the sa fety Tank is filled by linir.g 1here is a small heat exchanger Tank injection building, up var lo% valves in the and circluating gump which keeps west of the waste PAB.

Suction path to SIS the water > 4u'P, but these disposal building is automatically aligned not necessary following the loss o f AC.

Therefore, no eIcctrical power is needed.

Cit-MOV-52 4 u nside valor conta inment Control room 480V Energency MCCI Cat-MOV-523 In PAB next to charging Control room 480V Eniergency MCCl pumps a

Cil-MOV-525 inside vat,or containment Control room 400V HCCI Bus 1 ICV-222 In PAB Control room M

Y m

ma

T Tr.P.-C5 257-310 M undancy Following a loss of of fsite ac power, the centrol air ccupressors would not be available, since both compressors are powered fras non-energency 480-V busas. Ioss of both compressors eventually results in a canplete loss of control air to safe shutdown equipment. Operating Procedure 3002, "It 6s of Control Air Supply," defines the LT$edia'e operator actions required to place the plant in a hot shutdown condition.

Upon loss of of fsite ac power and control air supply, safe shutdown ccmponents are aff ected as follows:

o W-4 05 (auxiliary steam trip valve) will close. This valve supplies steam to the turbine-driven energency boiler feed pump and the la ge and small hoggers.

o

'K"/-2 0 0 (cooling water return CCW system) will open. This valve controls cceponent cooling water flow through the LPST ccoler and the y

SDC cooler.

o IC/-222 (letdown down line level control valve to LPST) will close.

This valve controls letdown flow rate to LPST.

o W-411 (atmospheric dump valve) will close. This valve provides a path for sensible and decay heat renoval.

o C/-1100A, B,C,D (bypass valves around feeJ. water blocking valves) will lock in position. These valves provide a bypass path around motor-operated feedwater blocking valves.

o C/-1100, CV-1200, CJ-1300, and C/-1400 (feedwater regulating valves) will lock in position. These valves control feedwater to t'..

steam generators.

o Wide and narrow range steam generator water level transmitters will indicate low steam generator level.

variable speed charging pumps will not control from the main control o

board.

(Cne charging pump centroller is locked in the full speed position.)

o Indications frem pneumatic level instruments will be erratic, including presaurizer pressure-Following a loss of centro. air, the operator can valve in energency nitrogen supply to the at:nospheric dump valve and the auxiliary steam trip i

M.... Franklin Researc.h Center s-52 a cm a.# m r l

TER-CS257-310 valve. In addition, the operator would control steam generator levels by using manual bypass valves. With erratic pneunatic steam generator lavels, the cperator must rely on the electrically transmitted steam generator level signals in the switchgear room, or tba electric le rel indication on or-inside the main control board can be used.

The service air system is another source of pressurized air. Service air is provided by one 514 scfm service air compressor with V-belt drive and 10,0-hp motor. The vertical, single-stage, double-acting, reciprocating, water-cool +5 compressor is of the lubricated type and is installed with intake filtne and air receiver. The 100-hp, 440-V motor is powered from non-emergency bus 4-1 and controlled by an air cireuit breaker with 125-V de control in the 440-V switchgear. Operation of the circuit breaker is controlled from a locally mounted 3-position switch. The type of automatic

$ontrol for this compressor duplicated that provided for the control air compressors.

The discharge from the service air receivers supplies a si.cgle header piping system which runs throughout the station. This system is interconnected with the control air systen.

The service air system can function as a backup to the control air system during normal operation.

Following a loss of of fsite power and/or a single failure, the system cannot be relied upon to fanction and therefore is not considered to be a safe shutdown system.

The control air system pro. ides pressarired air for necessary valve control functions within safe shutdown systems and pneenatic signals in assential instrumentation. Loss of the compressed air system will not prevent reaching a safe shutdown condition, but it is detrimental from the standpoint of causing numerous manual valve cperations and erroneous indication to the These additional acticns are beyond the ihnited operator actions operators.

that may result from a single failure.

Based on the abcve discussion, the staf f concludes that the control air system does not satisfy the functional require:ents of BTP RSB 5-1 in thar. a M

3-53 t.W Franklin,N rr h Center Researc 4 :>

a.

~

j

TER-CS257-310 reliable source of control air is not avaihble and significant operator action outside the centrol room is required to ef fect a safe shutdown. The staff will evaluate the significance of this in the SEP integrated assessnent of Yr.nkee Powe.

Locatien and Oeeration The staf f evaluated the equipment discussed above with respect to its 1ccation and operability during a loss of of fsite ac power. Table 3.2-9 shows the equipnent's location, the points from which it may be cperated, and its power supply.

EMERGE!CY POWER SYSTDi Task: Supply a reliable source of'ac power to run the necessary equipment.

t i

Discussion The three emergency' diesel generaters (EDGs) are eaco rated for cantinuous cperation at 500KVA, 480V, 0.8 pf, and 1800 rpm. 'The engines are fast-starting, V-16 (cylinders ), two-cycle, water-cooled engines that are directly coupled to an air-cooled synchronous generator.

Each engine has a closed, self-centained water cooling cycle and is started with a 125-V de cranking motor that is supplied with power frcm an independent battery.

Air for operatien of the engine and for exling the generater and engine radiator is obtained frcm coef int.tke vents. The cooling air exhausts to the outside atmosphere, and the engine exhaust is via a it.affler.

Each EDG has a 275-gallon fuel oil supply tank which contains enough fuel for 11.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months at full load. A 30,000-gallon fuel oil storage tank can supply any supply *.ank via gravity flow. The storage unk Technical.cpecification minimum (8,000 gallions) can supply enough fuel for all E:Gs at required load for more than 7 days. HI-IDW level in the three supply tanks is annunciated in the control recxn.

O EJ Franklin Research Center 3~54 Aw m aom w nm.aua

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tR TPBLE 3.2-9 il 4

E0llIPMENT IOCATION OPERATION POWER SilPPLY Two control air First floir of 1 mal control (1) 480V MCC 2 Bus I comgiressors tuchine building switch (2) 480V MCC 1 pus 1 One service air First floor of Imally mounted 480V station service as u,

ccepressor turbine building switch switchgear Bus Ser~t. 4-1 us

[

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f TER-C5257-310 Redundancv

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The expected electrical load (minimum) during a shutdown and cocidown following a less ef offsite ac power is given below:

1.

Charging pump 50 hp 2.

Service water pump 125 hp 3.

Component cooling pump 125 hp 4.

Shutdown cooling (or TIST) pump 60 ho Total 360 hp Since each EDG is rated at about 536 hp (400 kW), each EDG is sufficient to sapply the necessS_y electrical loads during the shu ;down and cooldown of the plant.

Also, the plant 480- and 2400-V electrical systems are designed such that if a particular pamp is unavailable, breakers :nay be repositioned so taat the redandant pump or component may be energized.

The EDGs are further evaluated in the resolution of SEP Topics VII-3 and VII-2 (electrical pcrtion).

Location and Coeration The staff evaluated the equipment discusred above with respect to its i

l location and operability during a loss of offsite ac power. lable 3.2-10 shows'the equipment's location, the points fres which it may be operated, and its power supply.

3.3 Saf e Shutdown Instrumentation Table 3.3-1 lists the instruments required to conduct a safe shutdown.

The list includes those instruments which provide information to the control rocm cperator frem which the proper cperation of all safe shutdcwn systems can be inferred. These instruments are the RCS pressure and temperatures, 1

i pressurirer level, and steam generator level. Improper trending of these parameters would lead the coerator to investigste the potential causes. Cther instruments listed in the table ptovide the operator with (1) a direct check on safe shutdown systen performance and (2) an indication of actual or f

Lupending degradation of systert performance. The list of instruments satisfies the requirement of 3TP RSB 5--I for safe shutdown. The DBE a

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1%ergency diesel Diesel generator Control rocan Start 81 125V DC Bus 1 generators building Start 82 125V DC Bus 2 Star t 93 Battery Dist.

Switchboard 3 i

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TABLE 3.3-1 SAFE S!!UTDOWN INSTRUMENTS CCHPONEllT/

[3 SYSTIM IllSTHUMt.NT I_NSTRUMENT IDCATION 4

7 }'

EN Steam Generator Steam Generator Level Control Hoom (Indication)

(LT&LI EM-1001, 1101, py 1201, 1301 (WR); EW-1003, 1103, 1203, 1303 (NR))

g[

Steam-Jenerator Preusure

[g (PIT-MS-403, 4, 5, 6) a Auxiliary Feed System Domineralized Water Storage Control Ibom (Indication and alarm)

Tank Level (LIT-405)

Auxiliary Feedwater Flow Control Ibom (Indication) 4 Chemical and Volume Charging Flow (FI&FT-2) m E

' Control System Ictdown Flow (FI&FT-1)

Shutdown Cooling SCS Flow (FI-204)

Local (Indication)

System 1

Component Cooling CCW Flow (FI&FT-201)

Control Hoom (Indication) j Water System

)

Cooling Water Supply l

Temperature (TI-222) l Service Water System SWS Flow Incal (Indication)

Main Coolant System Pressurizer Level (tlR)

Control Room (Indication)

(PR-LD-0) g 4

Pressurizer Pressure Control loom (Indication )

g (PR-PD-6, PR-IT-700) 3 Main Coolant System Control Ibom (Indicatior.)

4 (MC-PD-9) g Safety Injection SIT Level i

Tonk (SI-LT-1) l

4

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3 4

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a TABLE 3.3-1 (Continued) p

?.

'4 CCHPONElfr/

SYSTEM INSTHUHF:NT INSTRUMENT IOCATION, 1

Jrlinary Water PWST Level Storage Tank i

l us Diesel Generator Generator Output Control Roan (Indication)

E (voltage, cur rent, 4

io frequency)

.l neergency Power 480V ac buseas (status)

Control Roan (IniGicating Lights and l

System Voltmeter )

2400V ;sc buses (status)

Centrol Rocan (Ammeter and Voltmeter )

j 125V de taases (status) i i

}

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TER-CS257-310 evaluations, which in many cases are not based on the sane assumptions as this review, may determine that additional instrumentation is required to achieve ant uaintain a safe shutdown following a DBE.

The design of the instru-mencation and controls used for safe shutdown will be evaluated later in the electrical portion of the resolution of SEP Tcpic VII-3.

l l

A Tf[d Frank!in Research Gr.te B-60 a w w w r w we..

TER-C5257-310 4.

SPECIFIC RESIDUAL HEAT REMOVAL AND OUIER REQUIRD4ENTS OF BPANCH TICENICAL POSITION $~1 3 ranch Technical Position 5-1 contains the 'functicual requirements discussed in Section 3 and the detailed requirements applied to specific systens ne areas of operation. Each requirement is presented below along with a description of the Yankee Powe system or component applicable to the requirement.

4.1 RER Isolation Recuirements Recuirement The following shall be provided in the suctien side of the RHR system to isolaae it frca the RCS.

1.

Isolation shall be provided by at least two power-cperated valves in series. The valve positions shall be indicated in the control rocza.

2.

The valves shall have independent diverse interlocks to prevent the valves fece 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.

3.

The valves shall have independent divecse interlocks to protect against one or both valves being cpen during an RCS increase above the design pressure of the RER systen.

Eva luation 1.

The Yankeat Rowe shutdown cooling system fSCS) suction line has two power-cperated isolation valves which do not have position indicat!.on in the control recm.

2.

Neither of the two SCS suction valves nre provided with "open permissive" interlocks. The opening of these valves is administratively controlled. The controls for the valves are located l

in the primary auxiliary building (PAB). Key lock switches control each MCV, and the key is in the custody of the shif t supervisor.

';he two sucticn valves, FCV 572 and MCV 554, are powered fr'_m MCC 1,

[

Bus fl.

A failure of power supply will noc ef fect the position of these valves (either cpen-to-close or close-to-open).

I 3.

Neither of the two SCS suctiori valves ure provided with "autoclosure" interlocks. The SCS pressure is controlled by the RCS pressure and O

3-N

...; Franklin Researen Center l

m ar m vme,mu, l

~

i '

l TER-C5257-310 SCS (LPST) pump perfonnance when the two systems are connected. To ensure that the SCS is not overpressured, the RCS overpressure protection system, which includes the SCS relief valves, is provided. This is discussed further in Section 4.2.

The staff has concluded that the deviations regarding the independent diverse interlocks f or the RER isolation valves that prevent opening uritil pressure has decreased below RER design should be corrected. The staff's i

position on these deviations is given in Section 5.2.

The deviation from the BTP regarding lack of auccmatic closure for RER isolation valve's is acceptable because of the canbination of adminstrative controls and alar:ns provided on the RER system. These alarms provide additional assurance that the operator action required by procedure will be taken to shut the isolation valves when RCS pressure is increasing toward RER design pressure.

The staff has concluded that the deviations regarding RER isolatien valve

!position indication in the control room should be corrected. The staff's position on these deviations is given in Section 5.2.

Recuirement One of the following shall be provided on the discharge side of the RER systen to isolate it fran the RCS:

1.

The valves, position indicators, and interlocks described in Section 4.1.

2.

Cne or more eneck valves in series with a normally closed l

j power-operated valve. The power-cperated valve position shall be indicated in the control recra.

If the RER systen discharge line is used for an ECCS function, the power-operated valve is to be opened upon receipt of a safety injection signal once the reactor coolant pressure has decreased below the ECCS design pressure.

3.

Three check valves in series, or 4.

Two check valves in series, provided that there are design provisions to pennit periodic testing of the check valves for leaktightness and the testing is perfor:ned at least annually.

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TER-C5257-310 Evaluation The Yankee Rowe SCS has two meter-cperated isolation valves in series on the system discharge. The position of these valves is not indicated in the control rcom.

Like the two SCS suction MOVs d,iscussed in Section 4.1, neither SCS

~

discharge MOV control circuitry is provided with an "open permissive" or

" auto-closure" interlock. The opening / closing of these valves is ad:inistratively controlled. The controls for these valves are adjacent to the controls for the suction valves. Like the SCS suction MOV control

' switches, these are key lock switches with the key under the control of the shif t supervisor.

The two SCS discharge valves, MCV-551 and 553 are powered from MCC 1 Bus

1.

A f ailure of this power supply will not eff ect the position of these valves (either open-to-close or close-to-open).

4.2 Pressure Relief Recuirements - Overpressure Prctection Fecu irement To protect the RER system against accidental everpressuriration when it is in operation (not isolated from the RCS), pressure celief in the RER system shall be provided with relieving capacity in accordance with the ASME bailer

~

and Pressure Vessel Code. The most limiting pressure transient during the plant operating condition when the RER system is not isolated frem the RCS shall be considered when selecting the pressure relieving capacity of the RER sy st em. For example, during shutdown cooling in a PWR with no steam bubble in the pressurizer, inadvertent operation of an additional charging pump or in' advertent cpening of an DCCS ace==ulator valve should be censidered in selection of the design bases.

Evaluat ion All operating PWRs have been required to modify plant operating ;rocedures and install the necessary hardware to ensure rhat the RCS when in a cold aad shutdcan condition is not cverpressuri:ed. The RCS low temperature 4

B~03 L *nnidin Research Center 4 Ormesan of N F arwen moteame L

TER-C5 257-310 overpressure protection system (LTOPS) must be capable of mitigating the most limiting mass and energy input events. The LTOPS will also afford protection for the shutdown cooling system (SCS), which at Yankee Powe is the equivalent of the RER system.

The RCS and SCS can be connected whenever RCS temperature is below 330*F and the RCS pressure is below about 300 psig. There are no interlocks associated with the two suction or two discharge MOVs, and their position is under administrative control. The SCS design pressure is 425 psig, and the system has two spring-loaded safety valves set to cpen at 425 psig.'

The RCS low temperature overpressure protection system is designed to prevent exceeding the 10CFR50 Appendix G (isothermal curve) limit during the design basis mass and energy input events. The LTOPS utilires the two SCS safety valves and the pressurizer solenoid-operated relief valve (SORV) with a manually enabled low pressure setpoint of 500 psig. By procedure, the SORV is s' witched to the icw setpoint when the RCS pressure has decreased to below 450 psig.

The L': CPS, while being specifically designed to maintain the RCS pressure within the Appendix G limits, is available for overpressure p'rctection of the SCS.

Each credible mass and energy input event is listed in Table 4.2-1 along with the peak RCS (hence SCS) pressure.

In a September 14, 1979 (8] safety evaluation the staf f reviewed the Yankee Tcwe L': CPS, Information concerning mass and energy input events a well as additional discussion of the LTOPS equipment, its enployment, testing, and associated technical specifications are further discussed in the staf f's evaluation.

The SCS design limit of 440 psia is based on the pressure limit ice the bellows seals employed in certain system valves.

If the bellows failed, the I

va'1ve stem packing would be subject to system pressure, and even if the packing itself f ailed, the SCS would not experience total loss of function.

The governing standard fer the allcwable pressure on the pipes and other major components of the SCS is Wrican Standard ASA 331.1, 1955. This standard allows the imposed stress.cf 115 percent of design during 10 percent of the operating period and 120 percent of design during 1 percent of the operating O

3-64 l

d)Enklin Resear.ch Center A :>m at w. r I

l

j3 TAllLE 4.2-1 IllOPS ENERGY AND HASS ADDITION EVENTS

?y

8 3g; Peak L' LOPS Single l'ressur e Ileat Intwit Source RCS Temperature Lines of Defense Failure (psig)

Energy %dition Events Io ca

j Core decay heat T <300

F 1 SV1 + SORV 1 SV 515 and RCP (thermal)

All heaters T <300

F 2 SV + SORV 1 SV 470 3'

RCP startup THCS = 50* AT = 100*F 2 SVs + SORV 2 SVs 513 (Note 2)

TRCS = 100* AT = 100*F 2 SVs't SORV 2 SVs3 520 TECS = 100* AT = 100*F 2 SVs + SORV SORV 452 THCS = 150' AT = 100*F 2 SVs + SORV 2 SVs3 531 THCS = 200* AT = 100*F 2 SVs + SORV 2 SVs3 538 THCS = 100* AT = 150*F 2 SVs + SORV 2 SVs3 536 1.

Note la One SV is asstuned initially unavailable since an SCS MOV closure o

is assumed to inititate the event. The closure of an SCS 9

suction MOV makes the SCS suction side SV unavailable.

2.

Note 2: The AT indicated is the differential temperature between the steam generator secondary water and the coldest water anywhere b

J in the RCS.

o 3.

Note 3: The Licensee's analyses assumed only the availability of the SORV, -nd took no credit for the SCS SVs.

The staff has found no failure which would disable both SCS SVs.

TABLE 4.2-1 (Cont.)

lm ib

Mass

{$-

Peak Mass Input 1 5' Inpst Rate Single Pressure Sout.e 3d itCS Temperature Lines of Defense Failure (psig)_

ph Mass Addition Events S

-2 1 CCP 30 200* F(T<300* F 2 SVs + SORV 1 SV 425 4

1 CCP 30 T <200

F 2 SVs + SOltV 1 SV 425 1 IIPSIP 220 200

F <T <300

F 2 SVs + SOltV 1 SV 450 1 IIPSIP 220 T <200

F 2 SVs + SORV 1 SV 450 1 1.PSIP 1108) 200

F <r <3 00

F 2 SVs + SORY 1 SV 700 On 1 LPSIP4,5 1100 T <200

P 2 SVs + SORV 1 bv 700 I

1 Train 6 1100 20 0

F <T <3 0 0

F 2 SVs + SORV 1 SV 700 1

4.

liote 4: Only the mass addition froen one LPSIP la postulated in this' temperature band since all llPSIPs and LPSIPs are de-energized, but an operator error during testing of the LPSIP could result in pass addition.

4 a

5.

Note 5: The peak RCS pressure in this event is above the Appendix G g

(isothermal) limits.

A U

6.

Note 6: An ITCS train is canposed of one LPSIP and one llPSIP operating 3

in series. The mass additio,n from a single train of ECCS equipuen t is postulated since only one train is energized in g

this temperature band.

o TER-C5257-310 period.* The Licensee states that Stone and Webster (AE for Yankee Rowe) specifications for the SCS are based on these ASA B31.1 requirements. These specifications are given below.

System Temcerature Allowable Pressure **

300*F 680 psig 200*F 700 psig 100*F 720 psig Since these ihnits are not e.iteeded during any of the po *ulated transients given above, the staif concludes that the SCS piping and majer components are adequately protected for the LTCPS design bate transients. The relief protection used, however, is not in accordance with the ASME code since an active component (SORV) is utilired. The staff does not consider this a significant deviation and concludes that the overall SCS pressure relief requirements of BTP 5-1 are met.

This evaluation of SCS overpressure protection also applies to the low pressure serge tank (LPST) cooling loop since the LPST loop design is identical to the SCS.

Recuir enent Fluid ditcharged through the RER system pressure relief valves must be collected and contained such that a stuck-open relief valve will not:

1.

result in flooding of any safety-related equipment 2.

reduce the capability of the ECCS below that needed to micigate the consequences of a postulated LOCA 3.

result in a ronisolable situatien in which the water provided in the RCS to maintain the ccre in a safe condition is dis;harged outside of the containment.

ASA B31.1, 1955, paragraph 123 (b).

- It should be noted that these pressures are above the allewable pressures (at ccmparable temperatures) required by Appendix G (isethermal curve) for th e RCS.

D B-o,7 JCu Franklin Research Center A ca a s N rr.- m.

TER-C5257-310 Evaluation 1.

The SCS relief valves (2) can discharge to either the low pressure surge tank (LPST) or to the primary drain collecting tank (PDCT).

During SCS operation, the SCS relief valve discnarge is valved directly to the LPST.

A ccccnon 6-inch (CD) header directs relief discharge from several sources to two eductors under water. The LPST has a capacity of 750 ft3,.and a level control system keeps the tank about half full. The tank and water level control is designed to take three pressuri:er steam volumes before tank pressure reaches 75 psig. The LPST has six safety valves which relieve to a ecmmon he ader. The header has a rupture dise which opens at 25 psig and relieves directly to containment.

If one of the SCS relief valves stuck open, then approximately 101 gpm* would be lost out the RCS (and SCS) system. In about 29 minutes, the LPST would overflow out the open rupture disc.**

In this situation, the following alarms would alert the cperator: LPST level and pressure downstream of LPST safety valves.. Since there is no safety-related equipment in the contaiment sump or on the contaiment floor where the LPST safeties and rupture dise would relieve, no flooding of ECCS-related equipment would occur.

2.

The SCS is not used during either the injection or the recirculation phases following 'a LOCA. Therefore, a stuck-cpen SCS relief valve does not reduce the capability of the ICCS equipment.

3.

The SCS relief valves are outside contaiment but relieve to the LPST, which relieves back inside the vapor container (VC) ; therefore, on a stuck-open SCS relief, there is no net less of RCS or ECCS fluid.

Recuirenent If interlocks re provided to autenatically close the isolation valves when the RC pressure exceeds the RER system design pressure, adequate relief capacity shall be provided during the ti;ne period while tne valves are closing.

Evalu atien As discussed in Sections 4.1 and 4.2, the SCS isolation valves (two suction valves and two discharge valves) are not furnished with auto closure features. Therefore, this requirement is not applicable.

101 gpm at 465 psig (110% of setpoint pressure).

- The saf ety valves and rupture disc would Open in about 15 min.

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TER-CS257-310 4.3 Pumo Protection Recuirements Recuirement The design and operating procedures of an RER systen shall have provisions to prevent damage to the RHR system pumps due to overheating, cavitation, or loss of adequate pump suction fluid.

Evaluation There are no autcmatic trip or other features associated with the SCS or LPST cooling pumps that are designed to protect these pumps frcm overheating, cavitation, or loss of adequate pump suction fluid. The 480-V breakers supplying power to the SCS and LPST cooling pumps are equipped with the following protective devices:

l 1.

inverse time magnetic overcurrent trip (adjustable frce 60-150% of 100 amps coil rating) 2.

instantaneous trip (5-12 times the overcurrent coil rating of 100 anps).

These features are designed to protect the power supplies frem an equipment fault, but under certain circumstances (e.g, overheating), the trips may protect the pump motors. The Licensee has not evaluated these features I

i with respect to cavitation, overheating, or loss of suction fluid.

The following indications could alert the operator (s) to an abnormal situation in the SCS:

1 1.

MOV-554, -552, Pcaition Indication i

2.

MOV-551, -553, Position Indication 3.

LPST level 4.

LPST pressure 5.

SCS inlet tenperature 6.

SCS or LPST pump discharge pressure j

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7.

SCS or LPST cooler temperature (dischar ge) 8.

SCS discharge (to RCS) flow 9.

SCS or LPST cooler control valve (TICV 200) positicn.

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TER-C5257-310 4.4 Test Recuirenents Recuirement The isolation valve operability and interlock ci-cuits must be designed so as to permit on-line testing when cperating in the RER mode. Te st. ability shall meet the requirements of IEEE Standard 338 and Regulatory Guide 1.22.

The preoperational :nd initial startup test program shall be in conformance with Regulatory Guide 1.68.

"he programs for PWRs shall include tests with supporting analysis to (a) confirm that adequate mixing of berated water added prior to or during cooldown can be achieved under natural circulatien conditions and permit estimation of the times required to achieve suen mixing, and (b) confirn cat the cooldown under natural circulation c'nditions can be achieved within the limits specified in the emergency operating procedures. Ccraparison with performance of previously tested plants of similar design may be substituted for these tests.

Evaluation The procedure used to test the operability of SCS isolat,1cn valves

.TV-551 through -554 requires the stepping of the SCS pump prior to cycling the valves. Alternately, the operability of these valves could be checked by transferring SCS cooling requirenents to the low pressure surge tank cooling system ifeed and bleed). Since there are no "open permissive" interlocks associated wit' any of the four MCVs (two suction valves and two discharge valves), it is not necessary to bypass intericcks.

Yankee Powe has conducted plant cooldowns using RCS ne* ural circulstion, but has not performed any tests regarding flow measurement, cooldown rates, or coren mixing. However, the staff believes that, with the boric acid cencentrations used for shutdown, adequate boren mixing will occur unde; natural circulation flow.

di! Frank!in Research Center 3-70 4 hon of N Femen mon e

TER-C5257-310 4.5 ocerational Procedures Recuirement The cperational procedures for bringing the plant from normal operating power to cold shutdown shall be in conformanet with Rgulatcry Guide 1.33.

ter pressurized water reactors, the ' operational procedures shall include

+prific precedures and infocuation required for cooldown under natural circulation conditions.

Evaluation The Licensee has procedures to perform saf e shutdown operations including shutdown to hot standby, cporation at hot standby, hot shutdown, cperation at hot shutdown, and cold shutdown including long-term decay heat removal. The Licensee has also provided its operating etaff with procedures for shutting

!down the reactor cnd for decay heat removal under abnocnal and energency conditions. These procedures describe operator action in the event of loss of systen or parts of system. fune:tiens nonnally needed for shutdtwn and cooling the core. Procedures for the operation of individual systems used in safely i

shutting down the reactor are also included in the plant cperating procedures. These peccedures were reviewed and are in conformance with I

Regulatory Guide 1.33.

In adcition, Secticn B, " Procedures," of the Licensee's Technical Specifications assures establishnent of written procedures in accordance with NRC standards and Pegulatory Guides (including 1

Regulatory Guide 1.33).

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4.6 Auxiliary Feedwater Sucely l

Fecuirement l

The seisnic Category I water supply for the auxiliary feedwater system for a PWR Shall have sufficient inventory to pennit operation at hot shutdown for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , followed by cooldown to the conditions permitting operation of the RF.R system. The inventory needed for ccoldown shall be based on the longest cooldewn time with either only ensite er enly of fsite power available with an assu:ned single f ailure.

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The main ec.$ ant systen cooldewn rates ar.d aur.iliary feedwater supply inventories under varying conditions are discussed in Section 3.2 and Appendix B, Part 2.

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TER<5 257-310 5.

RESCLUTION OF SEP 'ICPICS The SEP topias associated with safe shutdown have been identified in the introduction to this assessnent. The following discussions evaluate the degree to which the safety objectives of these tcpics are fulfilled at the Yankee Rowe plant.

5.1

_T_opic V-10.B - RHR Systen Reliability The safety objective of this topic is to ensure reliable plant s.hutdown capability using safety-grade equipment subj ect to tae guidelines of SRP 5.4.7 l

and BTP RSB 5-1.

The Yankee Powe PWR systems have been coupared with the criteria of BTP 5-1, and the results of.these comparisons are discussed in Sections 3 and 4 of this assessment. Sectica 3 discusses the way the functional requirenents are met and Section 4 discusses the shutdown cooling systen (SCS), which performs the function identified in BTP RSB 5-1 as residual heat cernoval.

Redundancy to the SCS is provided by the low prersure surge tank (LPST) system. The LPST systen is physically arranged in parallel with the SCS. The couponents (pump and heat exchanger) of both the LPST system and SCS are identical and share a ccr. men suction and discharge line in the shutdown cooling modo.

Both the. suction and discharge lines are isolated by two meter-cperated valves in series. The staff finds this degree of redundancy acceptable; however, the following deviations exist whian could impair de reliability of the system:

1.

The SCS suction ar.d discharge meter-operated isolation valves do not have position indication in the control rocm. The valves are I

cperated fecm the primary auxiliary building (PAB) and connot be operated frcm the control recrn.

2.

There are no provisions to prevent damage to the SCS ptrnp or LPST systen cooling pump due to overheating, cavitaticn, or loss of adequate suction fluid.

3.

In order to cool the reacter cociant system to the SCS initiation point, significant cperator action mest be performed.:ran outside the control rocm.

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TER-CS 257-310 The first deviation as it relates to the potential overpressuriration of the SCS or RCS is addressed under Topic V-ll,4.

The first :.wo deviations also relate to interrupting the operation of the SCS while the plant is shut down and being main':sined at a temperature. equal to or less than 330*F and at a pressure less than 300 psig. The conseqwmes of an inadvertent valve closure or pump failure are that the cooldown would terminate and the plant could starc to heat up.

Installation of valve position indicators ar.d pump protective trips would alert the operator of the abnormal condition bv.t would not preclude it frcin occurring. Other plant parameters that are monitored centinously in the control room are available to indicate the status of the cooldown to the coerator. In the event that the cooldown has been terminat.ed dce to a pump failure, the redundant pump and heat exchanger frcin the LPST 2ystem can be put into service.

l Two modes of plant status must be considered wbon evaluating the overall eff ects of a los:s of the SCS function and the acceptability of th:=

deviations:

(1) plant shutdown with the temperature being maintained at less than 330*F and at scme pressure greater than atmospheric but less than 300 psio (2) the plant shut down and cooled down to lesc than 20'0*F, the reactor vessel head removed, and the system pressure at atmospheric.

In the first case, if the SCS were disabled due to 2 pump failure, a second pump, from the LP.ST system, would be available for concinued cooldown.

If the disruption of SCS were due to valve pecblems, an alternate method of maintaining the cooldown wculd have to be employed. Cne such method would be to let the plant heat up and to remove the heat generated through the steam generators (f eed to the steam generators can be obtained frcra a variety of sources). This provides an acceptable method in which to restore the heac renoval fecra the primary system. In the second case, as defined above, if shutdown cooling we.re interrupted due to valving failures, adequate cocling of the reactor could be acecrnplished by keeping the core covered with water.

Based on the discussions above, the staf f concludes that, al*ough j

deviations fccrn current licensing practice exist, the Yankee Powe SCS can reliably perform its cooldown functioni in the unlikely event of a pump or i

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e TER-CS 257-310 valve f ailure, acceptable alternatives exist to maintain the plant in a conditien which will not endanger the public health and safety.

The third deviation relates to the amount of cperator action required to establish shutdown cooling. Branch Technical Position 5-1 states that 'a limited amount of operator action fcou outside the control room 'is pennissible.

In the case of Yankee Powe, substantial effect is required of the operators from outside the control room to decrease main coolant temperature and pressure to a point where the SCS can be placed in operation.

Most of the equipment that requires manipulation for cooldown is located and centrolled from cutside the Yankee Rowe centrol rcam. The staf f evaluation shows the thne available before any operator action is necessary to be en the order of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or more, i.e., without any cperater action, from inside or outside the control room, the f acility can sustain itself with the water inventcry at hand.

' The amount of cperater action required is not canpatible with the intent of the topic criteria. The staff will consider the need for increased control room ooerability of cooldown systems during the integrated as.sessment.

l 5.2 Toeic V-11. A - Recuirements for Isolation of Hich and Low Pressu e

~

I Systems and "boic V-l'..B - RER Interlock Pecuirements The r,afety objective of these tcpics is to assure that adequate measures see taken to pectect low pressure systems connected to the primary system frc7 I

being subjected to excessive pressure, which could cause failurcs and in some plants could cause a LOCA outside contai=nent. The current criteria for RER isolation and pressure relief are discucsed in Sectiers 4.1 and 4.2.

The Yankee Rowe SCS suction and discharge (isclatico) valves do not have i

1 any cpen pecniserve interlocks or autanatic closure features, and valve position indication is not provided in the centrol room. This deviation involvec vielating a pressure boundary between a high pressure system (RCS) and a low pressure system (SCS or LPST). The interlock and at.comatic closure features are required whenever the RCS is at a pressure greater than the design pressure of the SCS or LPST (300 psig). The nost ihniting case is when i

the RCS is at cperating tenperature and pressure. The SCS/LPST is isolated l

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TER-CS257-310 from the RCS at both the suction and discharge sides by two key-locked motor-cperated valves in series. Ar inadvertent cpening of the pair of suct.cn or disch1rge valves could cause overpressuriration of the icw pressure system, which could cause a pipe or system failure, thereby creating a loss of coolant accident (LOCA) outside contairinent. Current criteria recuire open permissive interlocks, which prevent opening th'e valves when a specific pressure diff erential exists across the valves.

In lieu of the open permissive interlock, Yankee Rowe has key-operated valves, operatsd locally in the primary auxiliary building, with the keys maintained under administrative

, control. Due to the potential severity of SCS cverpressuriration, tr.a Licensee will be required to provide (1) interlocks tc prevent opening of SCS isolatien valves until the main coolant system pressure is below SCS design pressure and (2) valve position indication for the isolation valves in the centrol room.

s f

The SCS isolation valves do not have autanatic closure interlocks to close the valves during slow i. reaseG in RCS pressure. This is to prevent RCS pressuriration with any SCS LSolation valves in the open position. Rapid increases in RCS pressure are discuased in the Section 4.2 evaluation of the icw temperature overpressure protection (LTOP) system.

Some of these rapid pressure increases occur sufficiently fast that an autanatic closurs :..tericek would not respend in time to prevent overpressuriration of the SC3.

Hownerc the staff ocncluded that the LTOP provides acceptable SCS and LPST cooling loop pressure relief for these rapid transients. The staf f has determined that the installation of autanatic closure interlecks would not be desirable cince two of the three L':GP relief valves are on the SCS, and autanatic-isoluien of the 3CS fran the RCS would render the L' CP system incperable.

However,

'7 the SEP integrated assessment the staff will evaluate the potential need for additional measures, such as control roon valve indications, to prevent RCS startup and pressuriration with any SCS isolation valves in the cpen position.

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TIR-0$ 257-310 5.3 Tooic VII Systers Fequired for Saf e Shutdown The safety objectives of this tcpic are:

1.

to assure the design adequacy of the safe shutdown sytem to (a) initiate automatically the operation of appropriate systems, including the reactivity control systems, such that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences or postulated accidents, and (b) initiate the cperation of systems and ecmponents required to bring the plant to a saf e shutdown 2.

to assure that the required systems and equipment, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, are located at appropriate locations outside the control gcus and have a potential capability for subsequent cold shutdown of the reacter through the use of suitable procedures 3.

to assure that only safety grade equipment is required for a PWR plant to bring the reactor coolant system from a high pressure condition to a icw pressure cooling condition.

Safety objective 1(a) will be resolved in SEP Design Basis Event revicws. These reviews w'ill determine the need for autanatic, initiation of l

safe shutdown systems to mitigate the consequences of accidents and transients.

Objective 1(b) relates to centrol room availability of the centrol and l

instrumentation systems caeded to initiate the operation of safe shutdewn systems and assures that the centrol and instrumentation systems in the control rocm are capable of following tbg plant shutdown from its initiation to its conclusion at cold shutdown cend;tions; this does not apply to Yankee Rowe, since the entire cperation of shutdown cooling is performed outside the control room.

Saf ety objective 2 requires the capability to shut down to both hot shutdown and cold shutdown conditions using systems, instrumentation and controls located outside the centrol room. Yankee Rowe has procedures which identify several methods of tripping the plant and methods to cooldewn, provide adequate instruction for detecnining the operability and condition of the essential plant equipme nt, and indie ste the surveillance instrumentation and instructions needed for interpreting the information.

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j TER-C5257-310 The review team visited each designated cperator station 'and assessed the 3

capability,of the plant staf f to perform the necess?ry operations. The staff concludes that the plant can perform these shutdown cperations.

Conformance to safety objective 3 was not evaluated but will be ecmpleted in part under SEP Tepic III-1, " Classification of Structures, Components, and Systm s (Seismic and Quality)," and in part under Design Basis Event reviews.

5.4.Tosic X - Auxiliarv Feed System (AFS)

The safety objective for this topic is to assure that the AFS can provide ad~ equate 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 ASE 10-1.

The EBFP system and backup method w'ere compared with SRP 10.4.9 and BTP ASB 10-1 with the following conclusions:

i 1.

The Yankee Powe Nuclear 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 seinic design requirements, and earthquakes, tornadoes, and ficcds.

2.

The AFS conforms to General Design Criteria (GDC) 45

(" Inspection of Cooling Water Systems"' and GDC 46

(" Testing of Cooling Water Sy stems"). CDC 5 (" Sharing of Structures, Systus, and Components")

is not applicable.

l 3.

The Yankee Powe AFS is not autccatically initiated. Modifications to provide for autematic AFS initiation in accordance with I4ssons Learned Task Force reccz mendations are under staf f review.

4.

The Yankee Powe AFS dces not have capability to autcx:atically terminate feedwater flow to a depressurired steam generator and provide ilow to tae intact steam generator. This is acccmplished by local, manual valve cperation.

T.e ef fect of this deviation will be assessed in the main steam line break evaluation for the plant.

5.

In 1967, the Licensee made modifications to the Yankee Powe plant to prevent the occurrence of f eed system waterha:ener. The staff is contf nuing its evaluatien of feed syste waterhammer on a generic basis. SEP Tepic V-13, "Waterhanner," applies.

6.

The technical specificatiens for the AFS will be reevaluated against current requirements under SEP Tcpic XVI, " Technical Specifications."

4 3~ 0 SOl Franklin Research Center A Onamen of The Franen moeue

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TER-CS257-310 6.

REir5Er.CES Staff Discussion ci Fifteen Technical Issues in Attac. tent to November 3,1976.vecc3rntn f.rcm Dit2ctor, NRR, to NRR Staf f, NUREG-0138, November 1976.

2.

Appendix A to Part 50 of tne Cdde of Federal Regulations, Title 10.

3.

L. H. Heider (YAEC, Letter to Office of Nuc1 ar Reactor Pegulation (NRC) 9

Subject:

Core XV Refueling.

26 March 1981 4.

E. McFanna (NRC)

Telephone Ccnversat ice with B. Jcnes (YAIC)

Subj ect:

Saf e Shutt.own for Yankee Powe.

27.varch 1981 S.

D. E. Moody (YAEC)

I Letter to D. G. Eisenhut (NRC)

Subj ect: NRC Pequirements foi iuxiliary Feedwater Systems at Yankee Powe Nuclear Power Station.

21 December 1980 6.

J. A. Kay (YAEC)

Letter to D. L. Ziemann (NRC)

Su bj ec t: Resolution of DiI " Category A" 7_mplementation Audit Outstanding 7tems.

9 April 1980 7.

L. H. Heider (YAEC)

Letter to H. R. Denton (NRC)

Su bj ec t: Compliance with Appendix R to 10CFRSO.

19 March 19 81 8.

D. L. tiemann (NRC)

Letter to R. E. Grece (YAEC) l Subj ect: Aw.7dment No. 59 to Facility Operating License No. DPR-3 for the Yankee Nuclear Power Station (Yankee Powe).

14 Jcptember 1979.

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TER-CS 257-310 APPENDIX B, PART 2 SAFE SHinDOWN WATER REQUIRDENTS Introduction Standard Review phn (SRP) 5.4.7, " Residual Heat Removal (RFR) System, "

and Branch Technical Position (BTP) ESB 5-1, Rev.1, " Design Requirements of the Residual Heat Renoval System," are the current criteria used in the Systenatic Evaluation Program (SEP) evaluation of systems required for safe shutdown. BTP RSB

."-l Section A.4 states that the safe shutdown systen shall be capable of bringing the reactor to a ccid shutdown condition, with only offsite or ensite pcwer avdilable, within a reasenable period of thne following shutdown, assuming the most limiting single failure. BTP RSB 5 *z Section G, which applies specifically to the amount of auxiliary feed system

{AFS) water of a pressuri:ed water reactor available for steam generator feeding, requires the seismic Category I water supply for the AFS to have sufficient inventory to permit operation at hot shutdown for at leact four hours, followed by cooldewn to the conditions permitti, cperation of the RER sy st en.

The inventory needed for cooldown shall be hased en "the longest cooldcwn tbne needed with either only onsite ce cely offsite power available with a'n assumed single f ailure. 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 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

For a re&ctor plant cooldewn, water is the medium for transfer of heat from the plant to the environs. Two modes of heat removal are available. The l

fi;ct mcde 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 deminerali:ed " pure" water stored onsite and, therefore, is 15nited in quantity. The systems designed to use this mode of heat removal (boiloff) ace the steam generators for a pressuri:ed water reactor (PWR) and the emergency (isolation) condenser for a boiling water reactor (EWR). The second heat renoval mode (blowdown) involves the use of power-cperated relief valves to remove heat in the form of steam energy directly frem the reactor coolant system. Since at is not acceptable to vent the reactor coolant system directly to the atmosphere, the steam is typically vented to the centainment O

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TER-C5257-310 building, frcra which contair:nent cooling water systems transfer the heat to an ultimate heat sink - usually a river, lake, or ocean. When the blowdown mode is used, reactor coolant system makeup water must be continuously supplied to keep the rtacter core covered with coolant to ccrapensate for the loss'of I

ccolant i.wemory. Systems employing the blowdown heat removal mode have been designed into or be xfitted ente, most BWRs. The ef ficacy of thu blowdown mode for PWRs has received increased staff attention since the Three Mile Island Unit 2 accident in March 1979. Additional studies are planned or in progress.

This evaluation of cooling water requirements for safe shutdown and cooldown is based on the use of the system identified in the SEP eview ci Safe Shutdown Systems which has been completed for each SEP f acility in accordance with SRP 5.4.7 and BTP RSB 5-1 criteria.

It should be noted that the SEP Design Basis Eveats (DBE) reviews, now in progress, may require the

cse of systems other than those evaluated in this report for reactor plant shutdown and cooldown. In those cases, the water requirements for safe shutdown will have to b, evaluated using the assumptions of the DBE review.

Discussion The requirtment in ETF RSB 5-1 and SRP 5.4.7 that a plant achieve cold shutdown conditiens within approximately 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months is based mainly on the desire to be able to activate the RER system and transfer the plant heat to an ultimate heat sing prior to the exhaustion of the limited amount of onsite-stored pure water available for the AFS of a PWR.

A s2 stained hot shutdown condition, with reactor coolant systens temperature and pressure in excess of RER initiation limits, requires centinued boiling of f of pure water to renove reactor core decay heat.

A SWR relying on the emergency condenser system for cooldown under similar conditions would also pctentially exhaust onsite-stored pure water.

If water stored ensite is Opleted, raw water, for example from a river, lake, or ocean, can usually be tapped to supply the boiloff systems. However, raw water can accelerate the corrosion of boileff system materials in the steam generater and emergency condenser tubes even if the water is fresh. Raw l

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I fresh water can cause caustic stress corrosion cracking of both stainless j

steel and incenel tubes in less thsn 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months through NaCH concentration.

Seawater can cause chloride strer.2 corrosion cracking of the tubes well with cne week. Plant cooldewn and depressurization would help ieduce the rate of j

tube cracking by reducing the stresses in the tube materials and would also i

reduce the leakage rate of reactor coolant through c :cks 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 ma'jority of operatirg plants, saf e shutdown was defined as hot shutdown.

Therefore, the design of the syates used to achieve a cold shutdown condition was determined by the reactor plant vendo: and was not necessarily based en saf ety concerns. Safe shutdnwn reviews have pointed out a dif ference in vender approach to system design for cold shutdown reflected in the Standard yechnical Specification definiticn of cold shutdcwn:

for a BWR, cold shutdown requires reacter ecolant temperature to be _y212*F; for a ?WR, the temperature is _<20 0

F.

This difference in cold shutdcwn temperatures requires additional systems for PWR cooling not needed for a BWR.

For example, a BWR could use isolation condenser alone to reach 212*F- (although the approach to the final l

temperatu:e would be asymptotic); but a PWR, in addition co the steam I

generators, must use RER and supporting systes to cool to 200*F.

Evaluaticn Table 1 provides plant-specific data and assumptiens used in the staf f calculation of safe shutdewn water requirements for the Yankee F. owe nuclear f

plant. Table 2 presents the results of the calculation.

Four hours af ter the reactor trip, the decay heat rate is 1.774 x 10 Stu/h, and the integrated heat over the 4-hcur period is 1.149 x 10 Stu.

To maintain a ecnstant reacter coolant temp erature of 538'? for the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the staff calculated that 12,270 gallons of auxiliary feedwater are required to remcve the integrated heat.

Folicwing a 4-hour delar period, steam is released through five steam vent paths:

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TER-C5257-310 1.

atmospheric dump valve (ADV) 2.

1-in vent 3.

large hogger 4.

snall bogger 5.

energency boiler feed pump (ESFP).

Assusing that the cooldown rate does not exceed the administrative Ihnit of 50*F/h and that no single f ailure event occurs, cooldown to 330*F requires an additional 44 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months and consumes an additional 59,700 gallons of makeup water. Final cooldown to 200*F is cccomplished by manually actuating the

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shutdown cooling system.

Based cn the staf f calculation, Yankee Rowe's existing steam vent paths do not have sufficient heat renoval capacity to achieve cold shutdown conditions within the Standard Review Plan 5.4.7 requirement of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months assusing loss of offsite power and a single failure. However, sufficient wa.ter inventcry is available to cend"et the plant cooldown.

In a March 26, 1981 letter (3), the Licensee proposed changes to provide autcmatic quick closure of the four main steam line non-return valves. This modificatica necessitated the installation of a new steam supply line to the steas-driven emergency feedwater pump and installation of additional steam dump capacity. During a March 27, 1981 discussion (4), the Licensee indicated that an additional manually cperated dump valve would be installed on each steam line upstream of the non-return valve. Each of these valves is to have the ability to remcve approximately 60,000 lbn/h. The staff repeated the safe shutdown water requirement calculation for these new steam vent paths assuming a single failure of cne at=osphecie dcup velve. Table 3 presents the results of the calculatien.

As in the previous case, the decay heat rate 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following the i

reactor trip is 1.774 x 10 Btu /h and the integrated heat over the 4-hour period is 1.149 x 10 Btu.

A total of 12,270 gallons of auxiliary fec3 vater is expended to renove the integrated heat. Steam is then vented through the three atmospheric dump valves. Assening that the cooldown rate does not exceed the administrative ILnit of 50*F/h, cocidown to 330'F requires an additional 4.35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months and consumes an additional 7,320 gallons of makeup 4

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1 water. Based on the abeve calculations, Yankee Bowe's preposeo stema vent l

paths have sufficient capacity to coiduct a plant cooldewn in accordance with i

BTP RSB 5-1.

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TER-C5257-310 TaCle 3.2-1 PLANT-SPECIFIC DATA AND CCCLDOWN ASSL?iPTICNS Plant-Seecfic Data Plant Yankee Bowe Power 600 MWt Initial RCS Temperature 538'F Secondary Makeup Water Temperature 100*F Desineralired Water Cnsite 85,000 gal Existing Atmospheric Dump valve 1.316 x 107 Btu /h at 935 psig Heat Removal Capacity Proposed Atmospheric Dump Valve 2.025 x '.08 Btu /h at 935 psig Heat Removal Capacity - 3 Valves Large and Ssall Hogger Heat 6.415 x 106 Btu /h at 935 psig Removal Capacity One Inch Vent Heat Removal 4.249 x 107 Btiu/h at 935 psig Capac ity Inergency Boiler Feed Pump Heat 2.388 x 106 Btu /h at 935 psig Pzuoval Capacity 1

I Cocidewn As?um=tions 1.

Peactor trips at t = 3.

2.

Decay heat is in acccedance with Draft ANS-5.1.

3.

Plant remains at hot shutdcwn for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to cooldown.

l 4.

Mass of water in the steam generates is constant.

5.

Administrative cooldewn rate of 30*F/h is not exceeded.

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TER-CS25-310 Table B.2-2 CALCULATION CF SAFE SHUTDOWN WA7'ER REQUIRE 4ENTS Plant:

Yankee Powe (Calculaticns performed using the existing steam vent paths).

Phase I (Peactor tr '.p to point at which decay heat generation equals the beat rernoval rate of the steam vent paths):

Time at which decay heat generation equMs heat :emoval rate:

9.6 min

~

Phase II (Four-hour delay prior to cooldown):

8 Decay heat generated prior to cooldown: 1.149 x 10 - Btu Feedwater expended prior to cooldown: 12,270 gal Phase III (Cooldowr F.ain Coolant Stearn Generator iime (h )

Temeerature (*F)

Pressure fesia)

Decav Heat Generated fBtu) 4.0 538.00 946.7 1.149 x 10 6.0 437.97 373.7 1.490 x 10 8.0 387.87 214.9 1.804'x 10 10.0 368.79 170.8 2.091 x 10 8

15.0 350.48 135.5 2.716 x 10 8

20.0 343.48 123.6 3.281 x 10 25.0 338.93 116.3 3.796 x 10 8

30.0 336.87

'.13. 2

4. 292 x 10 8

35.0 334.96 110.3 4.772 x 10 40.0 333.03 107.4 5.236 x 10 45.0 331.14 10 '..

i.685 x 10 4 13. 0 329.99

.03.0 5.946 x 10 Decay heac rate at t = 48.0 h 5.522 x 10 Btu /h Feedwater expended during ccoldewn to 330*F:

59,*iOO gal Tetal f eedwater exper.ded: 71,970 gal l

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TER-C5257-310 Taole B.2-3 CALCULAfICN OF SAFE SHUTTWN WATER REQUIREMENTS Plants Yankee Powe (Calculations performed using the proposed steau vent paths)

Phase I (Ikactor trip to point at which decay heat generation equals the heat removal rate of the p, reposed steam vent paths):

Time at which decay heat generation equals heat removal rate:

0.0 min Phase II (Four delay prior to cooldown):

Cecay heat generated prior to cooldown:

1.149 x 108 Stu FeeCwater expended prior to cooldown: 12,270 gal i

Phase III (Cov'Jown)

Time (h )

Tencerature (*F)

Pressure (psia)

Decav Heat Generated (Btu) 4.0 518.0 946.7 1.149 x 10 8

6.0 438.0 373.8 1.490 x 10 8

8.0 339.47 117.2 1.804 x 10 8

8.35 329.96 103.0 1.856 x 10 Decay heat rate at t = 8.35 n; 1.477 x 10 Btu /h Feedwater expanded during cooldcwn to 330*F: 7,320 gal Total feedwater expended: 19,590 gal I

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ML20005B896

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ML20005B896

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