ML20040A695

Forwards Reactor Sys Branch SER Input Re NSSS for B-SAR-205

ML20040A695
Person / Time
Site:	05000561
Issue date:	06/28/1977
From:	Ross D Office of Nuclear Reactor Regulation
To:	Vassallo D Office of Nuclear Reactor Regulation
Shared Package
ML111090060	List: ML19208A315 ML19208A325 ML19250B014 ML19250B445 ML19250B480 ML19250B482 ML19254C975 ML19254G119 ML19254G128 ML19255K396 ML19255K397 ML19257E274 ML19257F047 ML19257F050 ML19260B823 ML19260E912 ML19260E913 ML19260E914 ML19261C469 ML19262E467 ML19262F230 ML19263D836 ML19263F605 ML19263G068 ML19268C086 ML19268C087 ML19269F130 ML19271A085 ML19271A172 ML19273B051 ML19273D308 ML19274A409 ML19274E085 ML19275B428 ML19275B444 ML19275C327 ML19275C328 ML19275G904 ML19275G918 ML19275H308 ML19275H317 ML19275H339 ML19275H341 ML19275H375 ML19275H401 ML19275H405 ML19275H493 ML19275H655 ML19275H660 ML19276D594 ... further results
References
FOIA-80-515, FOIA-80-555 NUDOCS 8201210426
Download: ML20040A695 (54)

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UNITED STATES NUCLEAR REGULATORY COMMISSION f.

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JUN 2 81977 l

Docket No. 50-561 i

MEMORANDUM FOR:

Ocmenic B. Vassallo, Assistant Director for LWR's, DPM FROM:

D. F. Ross, Jr., Assistant Director for Reactor Safety, DSS

SUBJECT:

SAFETY EVALUATION REPORT - B-SAR-205 STANDARD NUCLEAR STEAM SYSTEM Plant Name:

B-SAR-205 Licensing Stage:

PDA Milestone No.:

24-21 Docket Nos.:

50-561~

Responsible Branch & Project Manager:

LWR-3, T. Cox Systems Safety Branch Involved:

Reactor Systems Branch Description of Review:

SER Input Review Status:

Complete The enclosure contains the Reactor Systems Branch safety evaluation report for the B-SAR-205 Standard Nuclear Steam System (Babcock & Wilcox)

We have reviewed B-SAR-205 throuah Amendment 14 and have provided in the safety evaluation report our review of Sections 3.5, 4.2.3, 5.2.2, 5.2. 7, 5.5.1, 5.5.2, 5.5.3, 5.5.7, 5.5.10, 5.5.13, 6.3, and 15.0.

1 Dose calculations for the main steam line break and the locked rotor l

accident are being performed by the Accident Analysis Branch.

Confirm-l ation of the doses resulting from these accidents is necessary to i

support the acceptance of these analyses.

It should also be noted that Section 5.4.7 discusses the acceptability of DHR overpressure protection.

Related to B&W's position on this item is the staff position on moderate and high energy pipe cracks l

Contact:

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S. Newberry, NRR x27911 g1 26 810403 NADDEN80-515 PDR

o 2-JUN 2 81977 D. B. Vassallo d

gp ch#h b e-outside containment.

A common power supply on each train also inter-faces with isolation requirements of ACSB and our acceptance of e*

this design assumes their concurrence.

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D. F. Ross, Jr., Assistant Director for Reactor Safety Division of Systems Safety

Enclosure:

SER Input

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ACRS Generic Issues Relevent to RSB cc:.

S. Hanauer V. Banaroya R. Heineman G. Lainas D. Ross T. Ippolito

0. Parr R. Fitzpatrick T. Cox D. Bunch G. Mazetis S. Israel S. Newberry G. Chipman J. Wing I

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3.5.1.2 Internally Generated Missiles (Inside Containment) i Criterhon 4 of the General Design Criteria requires that systems and components important to safety be protected against the effects I

i of missiles generated from within the containment.

The responsibility i

for protection of safety-related systems and components has been

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identified as outside the scope of B-SAR-205 (Balance of Plant).

Therefore, our review was limited to identifying (1) the sources j

of internally generated missiles and their characteristics, and 1

(2) the systems and components to be protected from missiles as a

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interface requirements for the applicant.

The selection of potential missiles for the design structures and ' barriers is the i

responsibility of an applicant referencing B-SAR-205 and must be further described in the applicant's SAR.

I We have reviewed the B-SAR-205 list of potential s.ources of internally l

generated missiles and their characteristics.

The potential for the reactor coolant pump and motor components to become missiles in i

the event of a rupture in the pump suction or discharge sections of the reactor coolant system piping is under generic study by the l

staff.

The Electrical Pcwer Research Institute (EPRI) and its contractors plan to complete a test program by late 1977 which is intended to provide data to assist in vendor predictions of reactor t

coolant pump overspeed.

(See Section 5.4.1.2) l I

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We conclude that considering: (1) the identification of safety equipment within the scope of B-SAR-205 and the safety equipment cited in the B-SAR-205 interface requirements of Section 3.5.6 which must be protected from missiles and (2) the description of post-ulated missiles generated from B-SAR-205 equipment, with the

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exception of reactor coolant pump components, B-SAR-205 conforms to the applicable Commission regulations, regulatory guides, staff technical positions' and industry standards.

Conformance to these requirements constitutes an acceptable basis for satisfying the applicable requirements of Criterion 4 of the General Design Criteri a, 4.6 Functional Desian of Reactivity Control System Reactivity control is provided. by control rod assemblies (CRA),

axial power shaping rod assemblies (APSRA), and burnable poison rod assemblies (BPRA). Additional control is provided by the addition of soluble baron by the makeup and purification system.

The control rod system will consist of A Control Rod Assemblies, 8 Axial Power Shaping Rod Assemblies with absorber material in the lower half of the rod only, and 116 Burnable Poison Rod Assemblies, all consisting of 24 full length rods.

Each control rod has a section of neutron absorber material, B C, and 4

is clad with austenitic stainless steel.

The icwer section of each APSR is made of a neutron absorber material, which is an alloy of silver-l

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indium-cadmium, and is clad with the same cold worked austenitic sta,infess s' teel.

Each BPRA is inserted into the fuel assembly guide tube and has a section of sintered Al 0 -8 C pellets to serve as 23 4 burnable poison.

The burnable poison is clad with cold-worked Zircaloy-4 tubing and Zircaloy-4 upper and lower end pieces.

The control rods will be used to compensate for reactivity changes due

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to variations in operating conditions of the reactor, such as power and temperature changes. _

The part length APSRA's will be used to maintain an axially balanced power distribution.

The burnable. poison assemblies are installed to control the reactivity cha'nge due to fuel burnup and fission product buildup and also to reduce the amount of soluble boron re-quired in the core.

The control rod worth is sufficient to provide the required 1% shut-down margin for a hot shutdown, while the soluble boron provided from the makeup and purification system must be used to provide the required margin for a cold shutdown.

For the B-SAR-205 accident analyses, a rod drop time of 1.9 seconds to two thirds insertion was used.

Rod worth data from operati,ng plants have shown the initial integral rod worth curves in B-SAR-205 to be non-conservative.

Considering this data, B&W submitted addi-tional analyses to the staff using new rod worth curves and a rod drop time of 1.7 seconds to two thirds insert 1;on.

Sensitivity s udies per-taining to this change in rod worth are included in B-SAR-205 and show

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that the results are acceptable.

Technical Specifications for

. applicant's ' referencing. the B-SAR-205 design must reflect this faster rod drop time..

The staff concludes that the designs of the reactivit.y control systems conform to all applicable regulations and are acceptable.

5.2.2 Overoressure Protection Protection of the primary systen against overpressurization during normal plant. operating conditions will be provided by pressure relief of the reactor coolant system from two pressurizer code safety valves and one pressurizer pilot-actuated relief valve.

The valves will discharge to the reactor coolant drain tank.

B&W has referenced BAW-10043, Supplement 1 as the basis for sizing their safety valves.

The report was submitted for the 3600 MWt plant and is not appropriate for the 3800 MWt p'lant.

The code safety valves are each rated to carry 500,000 lbm/hr at 2575 psig and are sized on the basis of the worst pressure transient (loss of feedwater).

The pilot-actuated relief valve has a capacity of 150,000 lbm/hr at 2295 psig.

The pilot-actuated relief valve is

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insialled so that the NSSS can acccmmodate a load rejection from full load without causing a reactor trip or actuation of the primary safety valves.

While more appropriate safety valve sizing analyses must be submitted at the FSAR stage for the 3800 MWt plant, examination of the B&W worst-case pressure transient (loss of feedwater) em e

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shows that sufficient margins exist to conclude that the B-SAR-205 design is acceptable for the preliminary design stage.

The peak pressure in the PSAR for the loss of feedwater event is 2740 psig, a value less than the 110% x design pressure criterion.

An applicant for a 3800 MWt license must submit a more applicable overpressure protection report required by the ASME code at the FSAR stage.

The most limiting transient with respect to main steam system integ-rity is the turbine trip from 102% x rated power.

An analysis of this event is present$ed in BAW-1004'3 Supp.1, Overpressure Protection for B&W PWR's with 205 Fuel Assemblies.

The B-SAR-205 application is for a 3800 MWt plant, thus all analyses must be based on 102%

power or 3876 MWt.

This topical report, BAW-10043 Supp.1, is based on 3780 MWt and the resulting peak main steam pressure for this event is approximately 1340 psig which is he'arjhe acce;itian~ce[cri-

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teria of 110".' secondary system' d6si"gh" pres'iu'rel(110 fill 235 s

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1358l psig) and,"therefore~ rEqisire7eanafis~is of ihis ' eve' t at'

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387631Wt prior 'tib granting UDA..

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With regard to the potential for pressure increases while a plant is shut down, there have been several reported incidents during startup and shutdown in which the limitations of Appendix G to 10 CFR Part 50 have been exceeded.

B&W has proposed a design using the relief valves in the Decay Heat Removal System to prevent an P

overpressurization event during low temperature operations.

For this design to function properly, it is necessary that the motor-I e

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operated isolation valves in the decay heat drop line be open.

B&W increased the interlock pressure setting (which shuts these valves to protect the lower pressure DHR system) from 475 psi to 675 psi to maintain these isolation valves open to provide a relief path for the reactor coolant system.

The staff has reviewed this

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design and c'an' sot _ find 'i~t accdptable for the 'fol,1,owing reaso,ns:

(1)

During shutdown plant operations with one DHR train isolated, as allowed by current Technic,al Specifications, the' single active failure of the remaining DHR motor-operated suction valve would remove the plant overpressurization protection at low temperatures.

D Ji Additional information is. required to identify and verify that the. worst-case overpressure event has been selected and analyzed.

The staff will require B-SAR-205 to commit to providing acceptable

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equipment design modifications which will satisfy at least the following requirements:

1.

Credit for coerater action.

No credit can be taken for operator action until 10 minutes after the operator is made aware that a transient is in progress.

2.

Sincie failure criteria.

The pressure protection system should be designed to protect the vessel, given any event initiating a pressure transient.

Redundant or diverse pressure protection systems will be considered as meeting the single failure criteria.

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3.

Testability.

Provisions for periodic testing of the over-pressure protection system (s) and components shall be provided.

The program of tests and frequency or schedule thereof will be selected to assure functional capability when required.

4.

Seismic desian and IEEE 279 criteria.

Ideally, the pressure protection system (s) should meet both seismic Category I and 1

IEEE 279 criteria.

The basic objective, however, is that the systeb(s) should not be vulnerable to an event which both causes a pressure transient and causes a failure of

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equipment needed to terminate the transient.

5.

Reliability.

The system (s) provided must not reduce the reliability of the emergency core cooling system or residual heat removal systems.

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' 5.2.5 Reactor Coolant pressure Boundary Leakace Detection R'eactor coolant pressure boundary leakage detection methods have been identified as outside the scope of B-SAR-205.

The diverse methods which must be employed for leakage detection to satisfy the require-ments of Regulatory Guide 1.45 and the General Design Criterion 30, Appendix A of 10 CFR part 50 must be provided by the applicant and described in his SAR.

5.4.7 Decay Heat Removal System The decay heat removal system will be designed to remove decay heat and sensible heat from the RCS and core during the latter stages of cooldown.

The system.also maintains the reactor coolant at refueling temperature, provides initial reactor coolant system circulation prior to startup, provides auxiliary spray to the pressurizer for complete depressurization after shutdown of the reactor coolant pumps, provides a means for filling and draining the fuel transfer canal, injects borated water from the borated water storage tank to the reactor vessel for emergency core cooling, and recirculates coolant from the containment sump thereafter.

The system will consist of two parallel flow trains each consisting of a decay heat removal heat exchanger, a decay heat removal pump, and the associated valves and instrumentation necessary for operational o

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control.

The ' suction lines of the system will be connected to the hot. leg o,f on,e reacEor cc,ol_ ant,lo,op,any.the return lines wi.ll.

be connected to the reactor vessel at the core flood nozzles.

The valve arrangement will be such that at all times the emergency core cooling system can inject into the reactor vessel should the i

need arise.

The decay heat removal system will be placed into

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operation approximately six hours after initiation of plant shut-down.

The system is designed to cool the reactor coolant system j

from 305 F to 140*F.in 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />,. assuming both trains ar'e operating.

Assuming the use of only one train, the plant can i

be shut down below 212*F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Inadvertent opening of the decay heat removal system while the reactor coolant system is at high pressure has been precluded through the use i

of interlocks which prevent opening of the isolation valves when the reactor coolant system pressure is' greater than 675 psig and automatically close-the isolation valves when the reactor coolant system pressure reaches 675 psig.

The power supply arrangement to these isolation valves is such that loss of one hower source during shutdown cooling will not allow isolation of one train from the control room.

For this reason, we find this design to be unaccept-able and must be modified prior to granting a PDA.

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The decay heat removal system must perform the function of achie-ving and maintaining a cold shutdown condition in spite of a single active component failure.

B-SAR-205 will be required to show that all functions of the DHRS needed to achieve and. maintain j

a cold shutdown condition are not vulnerable to a single failure and utilize only safety grade equipment.

For example, the rate l

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of cooldown is normally controlled with the DHR heat exchanger bypass valves. A malfunction may cause one of more of these valve to fail closed resulting in maximum flow being directed through the heat exchangers.

This could result in excessive l

cooldown rates.

Also, the staff is presently considering on a generic basis the question of whether the capability should be provided for transferring heat'from P

the reactor to the environment from normal reactor operating condi-tions to cold shutdown using only safety-grade systems, with only offsite or onsite power available, and assuming the most limiting single failure.

Ifi{.i?Jeterminedthatthislcapabilityshouldbe

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provided, the staff will require that the B-SAR-205 design and the designs of the balance-of-plant portions of applications referencing B-SAR-205 be modified accordingly.

It is the staff's judgement that such modifications are technically feasible and conclude that this matter can be left for post-preliminary design approval stage considera-tion.

In the interim, 3-SAR-205 must, as a minimum, demonstrate that l

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each, plant can remain in a hot shutdown condition for an extended pieriod' of time assuming the loss of. offsite power and using only safety-grade equipment.

Interface criteria for BOP' areas must be 4

included in B-SAR-205 to ensure this capability.

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5. 5' '

Comconent and Subsystem Desion

5. 5.1 Reactor Coolant Pumos e

5. 5.1.1 Descriotion The reactor coolant pumps will be sized to provide adequate core cooling flow and hence sufficient heat transfer to maintain acceptable margins within the parameters of operation.

The estimated design pump capacity is 108,500 gallons per minute.

2 Sufficient pump rotational inertia (100,000 lb-ft ) wili be provided by a flywheel to provide acceptable ~fTod,coastdown characteristics following~a los_s.o.f~pu'm) power such t'ha(tile rea'c' tor iieu' trail powef_.

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can' be redu'c.ed before DNB limits are exceeded.

The reactor coolant pumps will be single stage single suction, constant speed, vertical centrifugal pumps employing a sealing system consisting of mechanical seal assemblies arranged in a removable cartridge and a seal leakage chamber to prevent reactor coolant fluid leakage to the atmosphere.

r 5.5.1.2 Pumo Flywheel Intecrity Criterion 4 of the General Design Criteria requires that structures, systems, and components of nuclear power plants important to safety be protected against the effects of missiles that might result from equipment failures.

Because flywheels have large masses and rotate i

at speeds of about 1200 revolutions per minute during normal reactor i

operation, a loss of integrity could result in high energy missiles O

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and" excessive vibration of the reactor coolant pump assembly.

The safety consequences could be significant because of possible damage to the reactor coolant system, the containment, or the engineered safety features. Regulatory Guide 1.14 describes the integrity and design requirements for pump fl.ywheels under normal operating

onditions.

i The potential for the reactor coolant pump flywheel to become a missile in the event of a rupture in the pump suction or discharge sections of reactor coolant system piping is under generic study by the staff.

The Electrical Power Research Institute has contracted Combustion Engineering, Incorporated to perform a 1/5 scale reactor coolant pump research program.

The objective of the program is, in part, to obtain empirical data to substantiate or modify current-

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mathematical models used in predicting pump performance during a postulated loss-of-coolant accident.

The staff is following the development and performance of this program as well as other industry analytical and experimental programs an a generic basis.

We have determined that additional protective measures, such as prevention of excessive pump overspeed or limitation of potential conse-quences to safety-related equipment, are technically feasible.

If the results of the generic investigations of this matter indicate that additional protective measures are necessary to assure that an accept-able level of safety is maintained, we will require that they be implemented.

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5.5.2 Stedm Generators 5.5. 2.1 Descriotion The two steam generators are vertical, straight, tube-and-shell heat exchangers which produce super-heated steam at constant turbine throttle pressure over the power range.

Reactor coolant enters the upper hemispherical head through a single inlet nozzle, flows downward inside 16,000 Inconel tubes and discharges through two outlet nozzles in the lower hemispherical head.

The tube and tubesheet boundary are designed for reactor coolant design pressure and temperature to minimize the transfer of radio-active fluids generated within the core to the secondary system.

The steam generators provide a heat sink for the reactor coolant system and'are at a higher elevation than the core to improve natural circulation for decay heat removal.

To ensure proper operational control and overpressure protection for the main steam system, applicant's which reference B-SAR-205 must incorporate all interface requirements identifidd in the B-SAR-205 into their balance of plant design.

Incorporation of these inter-face requirements will be reviewed for each applicant referencing B-SAR-205.

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'5.5.3 Reactor Coolant piping Tile reactor coolant piping is arranged in two transport loops, each with two rea'ctor coolant pumps and one steam generator.

Included as part of the reactor coolant piping is the surge line, which transports reactor coolant between the pressurizer and the reactor outlet piping, and the spray line, which connects the reactor inlet piping and the pressurizer and which transports reactor coolant for use in pressurizer.

The reactor coolant piping design considers ali phases of plant operation for a 40-year life including pre-operational testing, nomal operation, abnomal operation and accident conditions.

5.5.,10 Pressurizer The pressurizer maintains the RCS pressure during steady state operation and limits pressurc changes during transients.

It contains a water volume sized to provide the capability of the system to experience a reactor trip and not uncover the low level sensors in the bottcm pressurizer head while maintaining th'e; pressure above the actuation point of the HPI system.

The steam volume is sized to provide the capability of the system to experience a turbine trip and not cover the level sensor in the upper head.

Reactor trip occurs-for levels less than 125 inches or greater than 380 inches.

Two ASME Code,Section III, safety valves are connected to the pressurizer to relieve system cverpressure.

Each Val'7e~pr~dv~ ides one-half the total relieving capacity.

An additional pilot-operated

relief valve is provided to limit the lifting frequency of the

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code relief valves.

The relief and safety valves discharge to the reactor coolant drain tank within the containment.

5.5.13 Safety and Relief Valves The pressurizer safety valves are bellows sealed, balanced, spring-loaded safety valves which are provided with a supplemental backpressure balancing piston for handling a bellows failure.

The pressurizer relief valve is an-electrically actuated, electrically controlled, pilot operated, pressure loaded, relief valve.

The combined capacity of the pressurizer safety valves is 1,000,000 lbm/hr., which was based on twice the maximum surge result-ing from the upset that produces the largest pressure transient (see subsection 5.2.2).

The maximum surge assumes no direct reactor l

trip, operator action or credit for actuation of the pressurizer relief valve or turbine bypass system.

The pressurizer safety j

valves prevent the reactor coolant system pressure from exceeding 110f of system design pressure.

The pressurizen power operated relief valve prevents undesirable lifting of the spring-loaded safety valves.

The staff evaluation of the B-SAR-205 overpressure protection is contained in Section 5.2.2.

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t 6.3 Emercency Core Cooling System 6.3.1 Design Bases Criterion 35 of the General Design Criteria and Section 50.46 of 10 CFR Part 50 require that an emergency core cooling system be provided which can perform its safety function assuming a single failure.

The B-SAR-205 emergency core cooling system will be designed to provide T-emergency core cooling during those postulated accident conditions

z where it is assumed that mechanical, failures occur in the' reactor coolant system piping resulting in loss of coolant. from the reactor j

vessel greater than the available coolant makeup capacity using normal operating equipment. The emergency core cooling system will i

also be available to protect against steam line break consequences.

The B-SAR-205 emergency core cooling system will be similar in design, size, and capacity to those of Pebble Springs, WNP 1 and 4 and Bellefahte. -

d The system design bases are to prevent core damage that would y

i interfere with adequate emergency core cooling and to minimize the amount of clad water reaction for any size break up to and in-cluding a double-ended rupture of the largest primary coolant line.

These requirements are to be met even with minimum engineered safety features available.

The emergency core cooling system will have the required number, diversity, reliability, and redundancy of components such that no single failure of accive emergency core cooling system equipment during the short term or no single failure of active or passive equipment during the long term of an accident will result in inadequate cooling of the reactor core.

Each i

of the proposed emergency core cooling system subsystems will be designed to function over a specific range of reactor coolant piping system break sizes, up to and including the cross-sectional area associated with a postulated double-ended break in the largest reactor coolant pipe (15.75 square feet is the double-ended area of the hot leg and 8.55 square feet is the double-ended area of the cold leg.)

The emergency core cooling system is also designed to mitigate the i

consequences of a main steam line break.

Following a steam line rupture cr spurious relief valve Iifting, the 'mergency core cooling e

system injects borated water into the reactor coolant system provid-ing makeup and the ultimate shutdown of the plant.

The range of steam line ruptures protected aoainst is up to and including the double-ended circumferential rupture of the largest pipe in the steam system.

6.3.2

System Design

In the event of a postulated design basis loss-of-coolant accident, mass and energy will be released from the postulated pipe break to the containment.

These releases will occur over a time period depending upon the particular loss-of-coolant accident that has been postulated.

Within this time period several phases may be considered to occur in terms of blowdown, refill, reflood, and post refloed phases. These are discussed separately below.

The blowdown phase of the accident is the time immediately following the occurrence of the postulated break during which most of the mass and energy contained in the reactor system, the primary coolant, and the metal and core stored energy will be released to the containment.

The refill phase is that time during which the lower reactor vessel plenum will be refilled to the bottcm of the core by the emergency

_19 core cooling system.

The reflood phase is that time during which the core will be recovered I

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by the emergency core cooling system and, for cold leg breaks, the time period during which most of the secondary energy will be removed from the steam generators.

The remaining energy in the secondary system, along with decay heat from the reactor core, will be released to the containment during the post-reflood period.

For hot leg breaks, i

the broken ' piping will provide a direct path for fluid frcm the core to travel directly into the containment without passing through the iteam generators.

Therefore, the secondary system energy will be removed at a much slower rate.

Following a postulated loss-of-coolant accident, the emergency core cooling system will operate initially in the passive core flooding tank mode and the active high pressure injection mode, and finally in the The emergen'y core cooling system will consist of recirculation' mode.

c two core flooding tanks, three high pressure injection pumps, and two low pressure injection pumps with provisions for recirculation of the borated coolant after the end of the injection phasm.

Various ccmbinations of these systems will assure core cooling for the complete range of postulated break si::es.

Each of the two' core flooding tanks will have a total volume of 1800 cubic feet with a normal volume of borated water of 1350 cubic feet i

and a corresponding volume of nitrogen gas of 450 cubic feet at a nomal operating pressure of 600 pounds per souare inch, gauge.

The minimum boric acid concentration will be 2270 parts per million.

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Each tank will be connected by a core flooding line directly to a reactor vessel core flooding nozzle.

Each line will contain an electric motor cperated valve and two in line check valves.

The motor-operated isolation ve.lve will also be provided with appro-3 priate incerlocks to assure that the valves will be open during power operation when availability of the flooding tanks is required.

Babcock and Wilcox states that:

1.

Controls will be provided for each motor-operated valve to open the^ valves when RC pressure equals or exceeds 750 pounds per square inch gauge.

The controls are part of the safety-related control and instrumentation.

2.

Visual indication of the position of the isolation valves (opened or closed) will be provided in the control rocu.

3.

Two separate alarms based on valve position are provided.

4.

In addition to valve opening, the final. Technical Specifications will pequire that power be removed from the valve operator once n

it is open and prior to taking the reactor critical.

The high pressure injection mode of operation wili consist of the operation of two centrifugal high pressure injection pumps, rated at 700 gallons per minute each at a design head of 2500 ft. A third high pressure injection pump is provided as an installed spare with electric power lined up to the same power supply as the pump not being used for normal makeup (standby pump).

One of the high pressure injection pumps cperates continuously during nomal power operation providing normal makeup water as part of the makeup and purification system.

Upon actuation-of an Engineered Safety.reature Actuation Signal, two high pressure injecticn pumps provide high pressure injection of

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boric acid solution into the Reactor Coolan,t System by taking suction on the Borated Water Storage Tank whose contents are maintained at 2270 parts per million boron concentration.

Low pressure injection will be provided by two residua.1 heat removal pumps rated at 5125 gallons per minute each at a design head of 385 feet which will initially take their suction from the Borated Water Storage Tank.

Upon actuation of the 1ow alahn'hoin.~ tie '507a~tedM

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Water Storage Tank,' suction will b'e transferred to the containment sump for the recirculation mode of operation.

The ECCS will then provide the long-term core cooling requirements by recirculating the spilled reactor coolant collected in the containment sump back to the reactor vessel through the core flooding line nozzles.

The changeover from low pressure injection to recirculation is accomplished manually from

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the control room with automatic backup available.

Should the break size be such as to maintain reactor coolant system pressure higher than the LPI pump head, the required flow is delivered by the HPIS by aligning the flow from the discharge of the low pressure pumps to the suction of the high pressure pumps.

This " piggy-back" alignment is accomplished manually from the control rocm.

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6. 3. 3, Desian Evaluation Babco'ck and Wilcox has identified the following motor operated valves in the proposed emergency core cooling system design which must not move from normal alignment during certain phases of the postulated loss-of-coolant accident. These valves and their functions are as follows:

Valve Function DR-VlA, VlB LPI Isolation Valves DH-V14A, V14B DH Cooler Bypass Valves DH-V3A, V3B DH Flow Control Valves i

CF-V2A, V2B CF Tank Bleed Valves CF-VlA, VlB CF Isolation Valves cF 156154 cF T* 9 GVed da O B&W has proposed locking out the power to these valves, which is acceptable to the staff.

The B-SAR-205 incorporates the same upper plenum vent valve assembly characteristic as previous B&W plants.

Based upon information provided by B&W (Reference 1), the staff has concluded that sufficient assurance exists that reactor internal vent valves are not inadvertently opening in operating reactors and that the possibility of a stuck open vent valve is acceptably low.

Reports to NRC shall specify any loose parts monitoring anomaly attributed to a vibrating vent valve or vent valve components.

B&W plants which incorporate acceptable valve surveillance requirements into their Station Technical Specifications need not include a vent valve flow penalty in their design and safety analyses.

i 4

5 9

P_erformance Evaluation 6. 3. 4, e

9 Tne ECCS criteria of Appendix X to 10 CFR Part 50 of the Commission's regulations require that:

1.

The calculated maximum fuel element cladding temperature shall not exceed 2200 F.

2.

The calculated total oxidation of the cladding shall nowhere exceed 0.17 times the total cladding thickness before oxidation.

3.

The calculated total amount of hydrogen generated from the chemical

~

reaction of theccladding withr. water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel excluding the cladding surrounding the plenum volume, were to react.

4.

Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5.

After any calculated successful initial operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for an extended period of time required by the long-lived radioactiv.ity remaining in the core.

The ECCS evaluation model and analysis results applicable to B-SAR-205 are contained in B&W topical reports BAW-10104, BAW-10102 and BAW-10074 The ECCS calculations were submitted in BAW-10102 using the B&W evaluation model described in BAW-10104.

BAW-10074 describes the evaluation of small break LOCAs.

l The applicability of these LOCA analyses is dependent upon the satisfaction of all in,terface requirements specified in B-SAR-205.

i,.

The review of each SAR referencing B-SAR-205 will confirm that the applicant meets all requirements in his balance of plant design.

BAW-10102 contains a generic break spectrum appropriate for B-SAR-205.

A spectrum of break sizes, configurations and locations were performed in accordance with the staff's guidelines set forth in " Minimum Requirements for ECCS Break Spectrum Submittals," dated April 25, 1975.

2 These analyses identified the worst break as the 8.55 ft double-ended

' break at the pump d_i.scharge.

The table below summarizes the results of the LOCA analyses in BAW-10102 which determine the allowable linear heat generation rate limits as a function of elevation in the core:

eeaK Liaccing Maximum Local Elevation LHGR Limit Temperature Oxidation (ft)

(Kw/ft)

(* F)

(%)

2 14.9 2097 4.1 4

16.2 2156 5.9 6

16.8 2126 5.3 8

15.3 2177 6.7 10 14.2 21 71 6.3 The maximum expected core-wide metal / water reaction based on nominal equilibrium operation is calculated to be 0.62%.

The maximum cal-culated core-wide metal / water reaction based on operation at the Technical Specification limits is 1.0%.

As shown above, the cal-culated values for the ceak cladding temperatures, local metal / water reaction and core-wide metal / water reaction do not exceed the allowable 10 CFR 46 limits of 2200*F,17%, and 1.0% re.soectively.

Sinc,e the approval of the ECCS model described in BAW-10104, several changes to the model have been submitted.

Sensitivity

~

studies performed by B&W indicate that the effect of these changes would not be significant; however, to have a referenceable worst break which is analyzed wholly in conformance with 10 CFR 50.46 Appendix K and to ensure that the break spectrum shape has not changed, additional analyses have been requested.

The staff requires that these analyses be submitted prior to granting a Construction e

Permit on the first plant referencing B-SAR-205.

BAW-10102 includes assumptions for the containment net-free volume, passive heat sinks, and operation of the containment heat removal systems with regard to the conservatism of the containment pressure analysis.

The staff'noted that data for the passive heat sinks were 1ess than the j

recommendations in Branch Technical Position, CSB 6-1.

B&W re-evaluated the containment pressure using the CSB 6-1 heat sink data.

This analysts resulted in a containment pressure slightly lower (less than 2 psi) than that used in BAW-10102 for the first 60 seconds after the accident I

and higher thereafter.

Calculations by B&W indicate that the effect on

[

t peak cladding temperature is less than a 2*F increase.

Based on these j

calculations, we conclude that the ECCS containment pressure analysis of BAW-10102 is in accordance with Appendix K to 10 CFR 50 of the l

Ccmmission's regulations.

For each plant referencing B-SAR-205 and thus BAW-10102, we will require a comparison of the significant containment parameters with those used in SAW-10102.

At the operating license stage of review for each plant referencing B-SAR-205, we will require a comparison of the containment passive heat sink assumptions used in this analysis to those that exist in the plant.

I

~

Appendix X to 10 CFR 50 of the Commission's regulations also requires that the combination of emergency core cooling subsystems to be assumed operative shall be available assu...ing the most limiting single active ~

failure.

Review of the most damaging single failure of ECCS equia-ment was conducted.

Consideration must also be given to the possibility that manual valves might be left in the wrong position and remain undetected before an accident occurs.

Appropriate administrative procedures or position indication are examples of methods to be considered to minimize this possibility.

Each plant referencing B-SAR-205 will be reviewed to insure that these methods are properly implemented.

The ECCS must also retain its capability to cool the core assuming ECCS degradation during the long ter;n, recirculation cooling phase following the accident.

The staff's review of.this area includes the affect of lea; age and leakage paths on auxiliary systems (e.g., pumps and control systems), the ability'of the operator to detect leakage in sufficient time to prevent further damage as a result of flooding, spray impingement, etc., and the. ability of the operator to isolate the leakage.

For this phase of the accident, the analysis for B-SAR-205 consists of considering possible mechanical failures such as pump seal and valve packing failures.

The largest sudden leak potential is identified as the sudden failure of an LPI pump shaft seal.

B&W assumed a maximum leakage rate of

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3.0 g'pm based on the assumptions that the pump carbon insert does not exist and that no pressure losses take place along the leakage path.

Since equipment separation, leakage detection and other aspects of long term cooling are in the balance of plant scope, a review of this area will be conducted for each plant referencing the B-SAR-205 design.

Confirmation of the validity of the leakage rate assumed will be made on each as-built design.

To allow an operating configuration with less than four reactor coolant pumps on the line, the staff requires an analysis of the predicted consequences of a LOCA occurring during the proposed partial loop 4

t operating modes.

B&W has indicated that this matter will be addressed at the time application is made for an operating license. This position is acceptable to the staff.

We have also reviewed the proposed procedures and the system design foi-preventing excessive boric acid buildup in the. reactor vessel during the post loss-of-coolant accident long-term cooling period (presented in BAW-10102 as a generic analysis).

B&W has stated in BAW-10102 that a description of a baron dilution system for each specific plant will be provided on a case-by-case basis, such that each applicant which references ;

B-SAR-205 must provide a specific evaluation for his plant.

This position 1

is acceptable to the staff since the administrative procedures are largely BOP oriented and past experience has shown that design modi-fications are not major.

The staff requires that all remotel.y operated equipment located within j

i the containment that must function after a LOCA must be installed above we

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28 I

' the predicted post-LOCA flood level inside the containment.

Since the containment structure is within the balance of plant scope, it will be necessary for the staff to review each plant referencing B-SAR-205 for conformance to this requirement.

Babcock and Wilcox has specified that the following valves within the B-SAR-205 scope of design must be installed to meet this requirement:

System Valve No.

Valve Function Makeup and V2A, V2B Letdown isolati' n o

Purification V39A, B, C, D Seal return isolation Decay Heat VilA, VilB Drop line isolation Removal V12A,.V12B Drop line isolation V17A, V17B Boron Precipitation Prevention Additional information is required by the staff to evaluate the consequences and necessary actions after a break in a high pressure injection line.

Since the break location may not be apparent, a complete description of indications and operator actions is necessary for evaluation.

To provide adequate cooling for the core during the long-term re-circulation phase, the required net positive suction head (NPSH) must be available.

Since the containment sump is in the applicant's scope of design, his SAR must provide the calculations to show adequate level in the recirculation sump and the required NPSH.

To minimize the potential for water harmer, we will also require that venting provisions be described;in the final ~ design application for the emergency core cooling system.

S

' 6. 3.5' Tests and Inscections Th'e operability of the emergency core cooling system can be demonstrated by subjecting all components to preoperational tests, periodic testing, and in-service testing and inspections.

Each individual applicant referencing the B-SAR-205 design must describe his complete position on the preoperational testing described in Regulatory Guide 1.68, "Preoperational and Initial Startup Test Programs for Water-Cooled Power Reactors," Regulatory Guide 1.1, " Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps, and 1.79, "Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water Reactors."

The B-SAR-205 state.s that the following testing will be done on B&W ECC systems:

1.

HPI System - Each train must demonstrate an injection flow capability to the Reactor Coolant System.

Alignment of the HPI pumps taking suction from the LPI pumps will be verified.

2.

LPI System - Each train must demonstrate an injection flow capability to the Reactor Coolant Systerf.

Recirculation capability frem the sump will be verified for each, plant referencing B-SAR-205.

~

3.

Core Flooding - Core flooding tank flow will be verified on the

'l it prototype design to check the core flooding system and injection line to verify proper flow rate.

Subsequent units have a functional test to ensure that the lines are free of obstructions and that the core flooding line check valves and isolation valves operate correctly. CFT line resistances

1

^

shall be verified by B&W to be appropriate for each applicant 4.

Component Tests - Instrumentation per'formance, valve response timer under maximum differential pressures / temperatures and' pump capacity, discharge and flet Positive Suction Head are verified either during the safety feature and functional test or as part of a test for that individual component.

Continued periodic system and component tests will be conducted to maintain the assurance of proper ECCS performance.

The required tests and frequency will be provided in the applicant's Technical Specifications.

6.3.6 Conclusions 1.

Additional analyses are required to ensure that the break spectrum shape has not changed and to have a referenceable worst break which is analyzed wholly in conformance with 10 CFR 50.46 Appendix K.

The staff requires that these analyses be sbbmitted prior to granting a Construction Permit on the first p.lant referencing B-SAR-205.

F L

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l 2,

,The capability to cool the plant in the long term despite ECCS leakage due to seal degradation will be reviewed for each applicant referencing B-SAR-205, since the leakage detection systems and piping arrangements are in the balance of plant scope.

3.

Procedures for boron precipitation prevention will be reviewed I

on a plant specific basis.

4.

For each plant referencing B-SAR-205, a comparison of the containment heat sink assumptions used in..the analysis will be made to those that exist in the plant.

5.

Operation at less than full reactor coolant flow is not allowed since supporting LOCA analyses for such plant conditions have, not been submitted and approved.

6.

Additional information is required by the staff to evaluate the dohssqdEH665" and necessary actions after a break in a high pressure injection line.

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15.1 General ___

4 sabcock and Wilcox has performed safety analyses to evaluate the l'

capability of the nuclear steam supply system to withstand normal and abnormal operational transients and a broad spectrum of postulated accidents without undue risk to the health and safety of the public.

The postulated events have been. classified by Babcock and_Wil.cox._

with respect to evaluation criteria as followsf 1.

Condition I - Normal Operation and Operational Transients 2.

Condition II - Faults of Moderate Frequency 3.

Condition III - Infrequent Faults 4.

Condition IV - Limiting Faults.

Condition I events are those which are expected to occur in the course of normal power operation, refueling, maintenance, or maneuvering of the plant.

Condition I occurrences will be accommodated by suffi-cient design margin between any plant parameter and the value of that parameter which would require actuation of the reactor protec-tion system.

Condition I events will be handled by the reactor

~

control system which will automatically maintain prescribed condi-tions in the plant even under the most conservative set of reactivity parameters with respect to both system stability and transient

[

performance.

• The fundamental guicelines used by the staff are generally contained in the General Design Criteria, and appropriate Regulatory Guides.

When conflicts,between vendor classificatiors and the GDC exists, the GDC takes precedence.

i i

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-33.

Con,dition II events at worst will result in a reactor trip with the

' 1 ant being capable of return to operation.

Condition II events will not propagate to cause a more serious Condition III or IV event and are not expected to result in fuel rod failure or reactor coolant system.overpressurization.

Condition III events are very infrequent faults which will be

~

accommodated with the failure of only a small fraction of the fuel rods although sufffcient fuel dam ~ age might occur to preclude immediate resumption of operation.

For infrequent incidents, the plant should be designed to limit the release of radioactive material to assure that doses to personnel offsite are limited to values' which are _

small fraction of 10 CFR Part 100 guideline values.

A Condition III.

eventshall not generate a Condition IV fault or result in loss of function of the reactor coolant system or containment barriers.

1 Condition IV events are limiting design bases accidents which are not expected to occur, but are postulated because their consequences in-clude a potential for the release of significant: amounts of radio-active material.

System design for Condition IV events will prevent a fission product release to the environment which would result in

~

an undue risk to the health and safety of the public in excess of limits established in 10 CFR Part 100.

A Condition IV event is not to cause a consequential loss of required function of systems needed to mitigate the consequences of the accident, such as the emergency i

core ccoling system and the containment.

%%.ms.

The. Babcock and Wilcox classification of events analyzed is item-ized in Table 15-1 of this report.

15.2 Inout Parameters and Analytical Techniques for Accident and Transient Analyses 15.2.1 Inout Parameters As part of our review of the B-SAR-205 accident and transient analysis, we reviewed the assumptions and input parameters employed by Babcock and Wilcox in its analyses.

A df,scussion of the more significant assumptions and input parameters follows in this section.

Unless otherwise noted in this report, mathematical models and methods f

used by Babcock and Wilcox have been previously reviewed and found acceptable by the staff in conjunction with approved plants using

~

a Babcock and Wilcox nuclear steam supply system.

The uncertainties resulting from allowable operating bands and measure-ment uncertainties are reflected in the initial conditions or analyses trip set points that are equal to the Technical Specification set points plus maximum uncertainties.

Added conservatisms are stated in the analysis by use of clean or fouled (whichever is more conservative) steam generator inventory, minimum or maximum tank volumes, etc.

For DNBR calculations, transient pressures and temperatures are based on an initialization of offset conditions and are corrected by -45 pounds per square inch and +2 Fahrenheit, respectively,to account for control band and instrumentation errors.

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~+o one e Each applicant referencing B-SAR-205 must include these initial conditions and allowable operating ranges in their Technical Specifications. Their incorporation will be confirmed during the operating license review 'see Section 15.3).

All accident analyses in B-SAR-205 have been based on 102". of the rated core power level (3876 MWt) with the exception of the events

~ hic ~h-pioducEl'jje more. sever.e. consequences..at a. Iower.. power.. The ll

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B~SAR-205,acc,iden't ' analyses show'lha't the_. analyses.for the 3800 31Wt.

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olant ar'e " bounding for the 3600 MWt plant.

Except for the ECCS calculations, the analyses for B-SAR-205 are based on undensified fuel.

Penalties due to fuel densification will be considered at the FSAR stage, if appropriate.

The staff requires that the transients and accidents discussed in Chapter 15.0 be examined by each applicant referencing B-SAR-205 with regard to long-term effects.

The primary area of interest is with regard to the operator's role in achieving and maintaining stable conditions.

15.2.2 Analytical Technioues The folicwing analytical techniques used by Babcock and Wilcox in the B-SAR-205 accident and transient are under review by the staff.

SAW-10070 POWER TRAIll - General Hybrid Simulation for Reactor Coolant and Secondary System Transient Response ee

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-35 8,AW-10128 TRAP 2 2 FORTRAN Code to Simulate Once through Steam Generator Should modifications to these codes be required, the effect of these changes on the B-SAR-205 analyses must be censidered.

9 15.3 Technical Scecification Limits Oualified by Accident and Transient Analyses The results of accidents are sensitive to the value of many operating parameters which define conditions at the start of the transient and govern the response of the system model to the postulated accident condition.

Our review and approval of these analyses constitutes approval of the operating conditions and plant characteristics which have been found to be within the range that has been justified in the analyses.

The Technical Specifications must assure that operating conditions and trip setpoints are such that there is no potential for transients of more-severe consequences th'an those predicted in B-SAR-205.

Limiting core operating conditions which have been qualified by the accident analyses and which depend on parameters that are monitored directly or indirectly by sensor instruments are core power, maximum linear heat generation rate (LHGR), minimum departure frcm nucleate boiling ratio (DNBR), and reactor coolant average temcerature.

a

.---..-.e-.m

-37 These parameters relate to the staff's criteria for potential fuel damage and to the potential consequences of postulated accidents.

i In addition to directly monitored parameters, the axial and radial power distribution throughout the core are needed to evaluate the fuel damage potential due to high centerline temperature or clad burnout and the stored energy distribution.

The power distribution is dependent on the operating control rod configuration and crre physics and thermal-hydraulic parameters relating to core design and fuel loading and exposure distribution.

Limiting conditions of operation as specified in terms of the above parameters must be addressed by each applicant referencing B-SAR-205.

The average temperature used in the safety analysis should be in-cluded as well as the maximum linear heat generation rate and acceptable power distributions.

Also, the operating pressure and flow must be specified to be greater than or equal to the values used in the B-SAR-205 analysis.

In addition to the limiting conditions of operation, the Technical Specifications shall include the closure testing of isolation capability of turbine stop valves, feedwater isolation valves and main steam isolation valves.

The closure times are to reflect a consideration of the values assumed in chapter 15.0 analyses.

er:

y

The Technical Specifications must identify the administrative controls required and the system affected to prevent criticality during a refueling baron dilution transient.

As discussed in section 15.4.7 of this report, each applicant must show that, during refueling operations, 30 minutes exist between the indication (i.e., alarm)-

of a baron dilution and criticality.

To minimize the potential for lea-ving manual valves in the wrong position, applicant's referencing B-SAR-205 will be required to discuss and justify the planned administrative and design control placed on manual valves within systems intended to provide safety actions.

The Technical Specifications and accident analyses will be reviewed for each applicant referencing B-SAR-205 to assure that the pre-ceding areas have been appropriately. addressed.

15.4 Anticioated Transients A number of plant transients can be expected to occur with moderate frequency as a result of equipment malfunction or operator error in the course of refueling and power operation during the plant lifetime.

l Such transients meet the criteria of Condition II in the evaluation l

l and classification presented by Babcock and Wilcox.

l l

l We have reviewed the analyses submitted for these transients to 1

ascertain that the transients will not violate the following criteria:

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(1)

Pressure in the reactor coolant and main steam systems should be maintained below 110% of the design pressures (Section III of the ASME Code).

(2)

Clad integrity shall be maintained by ensuring that the min-imum departure from nucleate boiling ratio does not decrease below 1.30.

(3)

An incident of moderate frequency should not generate

-. a.more seribus plant cond'ition without other faults occurring independently.

The applicant has submitted analyses for these moderate frequency events to show that the integrity of the reactor coolant system pressure boundary has been maintained.

The worst primary side pressure transient was. identified as the complete loss of main feed-water from full power resulting in a peak pressure of 2740 psia in the reactor coolant system.

The worst secondary side pressure

~

transient was identified as the turbine trip without bypass resulting in a peak pressure of 1340 psig in the steam gen,erator.

The most limiting analysis with respect to core thermal margins is that for the feedwater temperature decrease which results in a minimum DNBR of 1.311.

B&W indicates that certain moderate frequency events in Chapter 15 may take credit for non-safety grade equipment.

Table 15.1 4 shows certain non-safety grade equipment was assumed to function.

This O

position does not appear to be consistent with Appendix 15C which depicts the systems required to mitigate the consequences of the Chapter 15 events and which do not show any non-safety grade systems required to perform a unique safety action.

Therefore, B&W will be requirbd to confirm and justify prior to a PDA that the overpressure and fuel damage criteria of the moderate frequency -

events in Chapter 15 would be met,with no credit assumed for unique actions performed by non-safety grade systems.

For example, the turbine trip transient should be analyzed at 102% power not taking credit for such systems as ICS or turbine bypass.

15.4.1 Increase in Heat Removal A number of plant transients can result in an unplanned increase in heat removal by the secondary system.

Those that might be expected

n to occur with moderate frequency can be caused.by feedwater system or pressure regulator malfunctions or the inadvertent opening _ of a steam generator safety or relief valve.

These t ansients have been reviewed and it was found that the most limiting in regard to core thermal margins is the feedwater temperature decrease event.

This transient was evaluated by the applicant using the POWER TRAIN computer code.

The results of the analysis of this event show that cladding integrity is maintained since the minimum departure from nucleate boiling ratio did not decrease below 1.311.

The B&W bounding

analysis for steam generator. safety or relief valve opentag or pressure regulator malfunctions assumes a 15% step increase in steam flow.

B-SAR-205 interface requirements require that applicants referencing B-SAR-205 must provide interlocks or some other means to limit turbine bypass (atmospheric and condenser dump) capacity to prevent incidents of moderate frequency from causing spurious opening of these valves which could result in an. increase of steam flow of more..than 15% rated. flow.

The acceptability of these interlocks will be determined during the review of the plants referencing B-SAR-205.

15.4.2 Decrease in Heat Removal Events A number of plant transients can result in an unplanned decrease in heat removal by the secondary system.

Our review of such events expected to occur with moderate frequency included turbine trip, t

loss of offsite power, loss of normal feedwater flow, and loss of condenser vacuum.

These transients were evalu'ated by Babcock and Wilcox using the digital computer codes POWER TRAIN and CADO.

It was found that the most limiting transient oIthi.s_ type with _. __

~

~

ruspectlto, core th'eiEal'~ margins was a turbine trip from.112% power.

The results of this analysis showed that cladding integrity was main-tained by ensuring that the minimum departure frem nucleate boiling ratio did not decrease below the acceptance criteria of 1.30 (minimum DNBR = 1.43 in the analysis).

The most limiting event of this type P

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l r,egarding reactor coolant system integrity was the complete loss of feedwater event with offsite power available throughout the transient.

The maximum pressure within the reactor coolant system reached 2740 psia, which is within the criterion of 110% of design pressure.

The most limiting transient with respect to main steam system integrity is the turbine trip without bypass from 102% x rated power.

An analysis of this event is presented in BAW-10043 Supp.1, Over-pressure Protection for B&W PWR's with 205 Fuel Assemblies.

The B-SAR-205 application is for a 3800 MWt plant, thus all analyses must be based on 102% power o~r 3876 MWt.

This topical report, BAW-10043 Supp.1, is based on 3780 MWt and the resulting peak main steam pressure for this ' event is approximately 1340 psig which is near the acceptance criteria of 110% secondary system design pressure (110% x 1235 - 1358 psig) and therefore, requires re-analysis of this event at 3876 MWt prior to granting a PDA.

15.4.3 Loss of Reactor Coolant Flow Several types of plant occurrences could result in an unplanned de-crease in reactor coolant flow rate. The most limiting transient of this type to occur with moderate frequency is the complete loss of reactor coolant flow with four pumps operating.

This event bounded the loss of offsite power transient.

This transient was evaluated by S&W using ea G

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+ - -

X"

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- t the PUMP, CADD and RADAR codes.

The results of the analysis showed i

that cladding integrity was maintained by ensuring that the minimum departure from nucleate boiling ratio did not decrease below 1.30, (minimum DNBR for analysis = 1.38), and that the maximum pressure within the reactor coolant and main steam systems did not exceed 110% of the design pressure.

i 15.4.4 Uncontrolled Rod Withdrawal, The possibilities for single failures of the reactor control system l

which could result in uncontrolled withdrawal of control rods under low power startup and power operation conditions have b'een reviewed.

i The scope of the review has included investigations of initial conditions and control rod reactivity worths, and the course of the resulting transients or steady-state conditions.

The Babcock and

~

Wilcox computer code CADD was used to determine the characteristics of these transients.

The most limiting rod withdrawal event with respect to core integrity was the withdrawal of all rods at full power.

The minimum' departure from nucleate boifing rat'io from this event remains greater than 1.30 (minimum DNBR for analysis =1.59).

The most limiting rod withdrawal transient with regard to reactor coolant system (RCS) integrity is a startup transient withdrawing a rod group worth 1.0 % AX/X, with two Reactor System Coolant Pumps operating.

Peak RCS pressure for this event is 2674 psia, which is belcw 110% system design pressure.

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7 15.4.5 Startup of an Inactive 1.000_

The startup of an inactive reactor coolant loop event was reviewed for B-SAR-205.

The most limiting transient is analyzed from the two reactor coolant pump made at 50% rated powc.a.

The Babcock and Wilcox code, CADD, was used to evaluate the transient response.

The minimum departure from nucleate boiling ratio does not decrease below 1.3 and Reactor Coolant System pressure remains below 110%

of design pressure, peaking at 2350 psia 17 seconds into the transient.

~

15.4.6 Inadvertant Initiation of ECCS~

The inadvertant operation of the Emergency Core Cooling System (ECCS) during power operation was reviewed.

Since the low pressure injection

' pump head is not sufficient to. inject water into the reactor coolant system during nomal operation, the analysis for this event concerns only initiation of the' high pressure injection system.

The boron concentration of the HPI water is sufficiently greater than the boron concentration at any point in core life (except refueling) that moderator dilution does not occur.

The worst-case dilution event t'

during refueling is discussed in Section 15.4.7.

The reactor will trip on high Reactor Coolant System pressure or hressurizer level to prevent system overpressurization.

Inadvertant initiation of ECCS during shutdown or startup is evaluated in subsection 5.2.2.

15.4.7 Chemical and Volume Control System Malfunction Various chemical and volume control system (CVCS) malfunctions which could lead to an unplanned boron dilution incident have been reviewed.

(

The staff notes that a bypass line to the makeup tank has been added to the B-SAR-205 design. Additional. information is required to i

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eea e show that the submitted analysis is appropriate, or to consider the additional boron dilution due to operation of the makeup tank bypass.

The concern is that the addition of this line to the plant design no lo~nger allows the dilution incident to be limited by the capacity of the makeup tank.

15.5 Postulated Accidents 15.5.1 Soectrum of Steam Svstem Pioinc Failures Inside and Outside of Containment TheanalysesandeIfectsofstea$1linebreakaccidentsinsideandout-side containment during various modes of operation and with or without offsite power have been. reviewed.

Review of the initial B-SAR-205 analysis required additional -infonnation to show that the full spectrum of breaks analyzed met all of the appropriate acceptance criteria.

The accident which resulted in the most severe conse-quences was the 42" double ended line break outside of containment with offsite power available assuming the single failure of the main steam isolation valve in that line.

Babcock and Wilcox states that they do not expect fuel cladding failures during the steam line break since less than 1.0". of the fuel rods experience a DNBR of less than 1.30 and since there is insignificant zirconium-water reaction in terms of the amount of wall thickness reduction that occurs as a result of the reaction (i.e.

00035 inch out of a total

[

wall thickness of.0235 inch).

The staff's current criteria for fuel i

cladding failure is a'DNBR 1.30. A dose calculation

• assuming that I cercent of the fuel rods fail shows that the worst case steam line

• LPM to see cover letter

(

break results in a dose which is an acceptable fraction of 10 CFR 100. cri teria.

The maximum pressure within the reactor coolant and m'ain steam systems did not exceed 110% of the design pressure.

The staff has requested additional information concerning once through steam generator flooding and a steam line break concurrent with steam generator flooding (Reference 3).

Because this matter includes a balance-of-plant interface, we will require that this material be submitted and reviewed prior to granting a CP on the first plant referencing B-SAR-205.

15.5.2 Feedwater System Pioina Breaks The analyses and effects of a spectrum of feedwater line breaks inside t

and outside containment, during various modes of operation, with or without offsite power have been reviewed.

Review of the initial B-SAR-205 analysis required additional information to show that the full spectrum of breaks analyzed met all of the appropriate acceptance criteria.

The feedwater line break which resulted in the most severe consequences with respect to the core, was the feed-water line rupture between the steam generator and the first up-stream check valve with loss of offsite pcwer at rupture.

The Babcock and Wilcox computer code CADD was used to determine the

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r characteristics of this accident.

The results of the analysis showed that no fuel damage occurred in that the minimum departure f

frem nucleate boiling ratio reached was 1.45.

The feedwater line A

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break upstream of the first upstream check valve is the worst case event analyzed in regards to overpressurization cf the reactor coolant system.

The relief capacity of the pressurization safety valves will be.eufficient to limit system pressure to 2740 psia which is less than 110% system design pressure.

15.5.3 Reactor Coolant Pumo Shaft Seizure / Break The analyses and effects of an instantaneous seizure of a rotor and on instantaneou,s. break of a s, haft of a reactor coolant pump during any allowed mode of operation have been reviewed.

The more limiting of these events is the pump seizure since no coastdown flow would be provided.

Babcock and Wilcox has classified this accident as a Condition IV event.

The staff considers it to be in a Condition III event which requires that the plant. be designed to limit the release of radioactive material to assure that doses to persons offsite are limited to values which are a small fraction of 10 CFR Part 100 guideline values.

The PUMP, CADD, and RADAR codes were used for this analysis which showed that 4.8% of the fuel rods experience minimum departure frem nucleate boiling ratio below 1.3 and that the peak clad temperature reached was 1325"F.

The analysis showed that the maximum pressure within the reactor coolant and main steam systems did not exceed 110% of the design pressures.

The staff concludes that the plant design is acceptable with regard to a possible seizure of a rotor or break of a shaft of a reactor coolant pump since the result dose is an accept-able fraction of 10 CFR 100.*

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15.5.4 Anticioated Transients Without Scram A number of plant transients can be affected by a failure of the scram system to function.

For a pressurized water reactor, the most l

important transients affected include loss of normal feedwater, loss of electrical load, inadvertent control rod withdrawal and loss of normal electric power.

In September 1973, we issued WASH 1270, " Technical Report on Antici.-

pated Transients Without Scram fo'r Water-Cooled Power Reactors" establishing acceptance criterie for anticipated transients without

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scram.

Babcock and Wilcox analyses for such transient; are discussed in BAW-10099. Rev. O.

On December 9,1975, we issued our staff Status Report which identified guidelines for further analyses, and in a staff letter of April 7,1976 we required B&W to provide analyses by June 30, 1976 and also to identify design changes to meet ATWS limits.

l Subsequently, B&W requested a delay for submittal of these analyses

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In December 1976, B&W provided partial ATWS analyses with a more complete analyses expected in May 1977.

The staff is continuing a generic review of this area of concern and the staff evaluation r

of the B&W analyses are expected to be published this year. The

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B-SAR-205 plant will be required to comply with the recommendations of the staff report.

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-G-I 15.5.5 Partial Looo Ooeration To allow operating configuration with less than four reactor coolant-pumps on the line (partial loop), the staff requires consideration of the predicted consequences of all accidents during the proposed partial loop operating mode.

Operation of the plant at the reduced power levels described in Chapter 15 for the associated pump ccmbina-tions is not allowed until additional LOCA analyses are submitted for these conditions. (see sectio'n 6.3).

At that time the staff will re-evaluate overall plant response during operation with less than four reactor coolant pumps.

15.6 Conclusions On the basis of our review of the B-SAR-205 accident and transient analysis, the following items must be resolved prior to preliminary design approval:

(1)

B&W must confirm that the overpressure and fuel damage criteria of the moderate frequency events in Chapter 15 would be' met with no credit assumed for unique actions p'erformed by non safety grade systems.

(2)

The bounding decrease in heat removal event with respect to

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main steam pressure, turbine trip, must be analyzed at 102P.

power (3876 MWt).

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(3)

B&W must provide additional information concerning the makeup tank bypass line in the chemical and volume control system to show that this system meets the staff requirements for dilution events.

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50-Table 15-1 B-SAR-205 Categories of Typical Transients and Faults Condition I Reactor startup Reactor shutdown Refueling operations Power operation i

Condition II - Faults o'f Moderate Frequency Uncontrolled Control Rod Group Withdrawal from Subcritical Condition Uncontrolled Contrni Rod Withdrawal at Pcwer Control Rod Misoperation Chemical and Volume Control System Malfunction Loss of Forced Reactor Coolant Flow (pump coastdown)

Startup of an Inactive Reactor Coolant loop Loss of External Electrical Load and/or Turbine Trip Loss of Normal Feedwater Loss of All a-c (off-site) Power to Station Auxiliaries Excessive Heat Removal j{

l Failure of Rc]ulating Instrumentation Internal and External Events (applicants' SAR)

Inadvertant Operation of ECCS during Power Operation Condition III - Infrequent Faults Fuel Misloading 1

Waste Gas Decay Tank Rupture 6.

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-50a-Table 15-1 (cont'd)

Condition IV - Cimi_ti[nj Fa~ults' [,

t Loss of Coolant Accident Steam Line Break Steam Generator Tube Rupture Rod Ejection Break in Instrument Lines or RC System Lines Penetrating Containment Fuel Handling Accident Reactor Coolant Pump Locked Rotor / Shaft Break Reactor Head Drop Accident note:

All other transients discussed in Chapter 15 of B-SAR-205 are shown to be bounded by the above analyzed events l

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References i

1.

Ltr. from A. Schwencer to Mr. Kenneth E. Suhrke, dated November 19, 1975.

2.

Ltr. from S. Varga to Mr. ' James J. Taylor, dated May 10, 1977.

3.

Memo for D. B. Vassallo fm. G. R. Mazetis, delivered to B&W March 15, 1977.

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ACRS GENERIC ISSUES RELEVANT TO REACTOR SYSTEMS l

BRANCH REVIEW 0F B-SAR-205 The following issues are categorized as Group I (resolved).

Reference to the applicable SER section is made below:

ISSUE SER Section I.1 NPSH for ECCS pumps 6.3.4 I.17 Operation of Reactor With less than All Lovps in Service -

6.3.4, 15.5;5 I.A.2 Primary System Detection and Location of Leaks 5.2.5 I.A.4 Anticipated Transients Without Scram 15.5.4 I.A.5 ECCS Capability of Current &

Older Plants 6.3 The following issues addressed as Group II items are under study, a Resolution is not considered necessary for the pr.eliminary design ap-

!1 proval requested by Babcock and Wilcox.

II.6 Common Mode Failures II.7 Behavior of Fuel under Abnormal Conditions II.10 ECCS Capability for Future Plants II.A.4 PWR Pump Overspeed During a LOCA II.A.5 Isolation of Low Pressure from High Pressure Systems II.C.1 Locking Out of ECCS Pcwer Operated Valves II.C.5 Waterhammer

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f UrJITED STATES '

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WASHINGTON, D. C. 20565

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JUL 2 21977 Docket No. STN 50-561 VENDOR: Babcock & Wilcox Company (B&W)

SUMMARY

OF MEETING TO DISCUSS B&W RESPONSES TO STAFF POSITIONS ON TWO OUTSTANDING ISSUES IN BSAR-205 REVIEW On July 12. 1977 representatives of B&W and our staff met in Bethesda to discuss B&W's responses to our positions on decay heat removal system isolation and overpressure protection of the reactor coolant system.

Our positions are described in Sections 5.2.2 and 5.4.3 of the Report to the ACRS issued on July 8.1977 and are the subject of outstanding

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issues which must be resolved prior to our decision on issuance of a Preliminary Design Approval for BSAR-205. Enclosure 1 is an attendance list.

B&W first presented their approach to OHRS isolation in the event of a pipe failure in the DHRS outside of containment. Their position is included as Enclosure 2 of this summary. B&W would establish an interface requirement in BSAR-205 that would require a referencing applicant to provide power to the required valves in such a way.that no single active component.

failure would result in loss of power to either of the two series valves in either of the two OHR trains. B&W states that the. failure of an electrica' bus by a 3-phase.. fault should not be assumed (as a single active failure) in conjunction with the DHRS pipe failure outside containment. They contend that the concurrent failure of both the piping and the electrical bus....

are events for which the combined probability is too low to warrant design changes to accommodate such a postulated event. Staff members indicated their intent to. require the assumption of bus failure, and also indicated that the probability values which B&W selected from NASH-1400 may not be conservatively applicable to the types of failures involved in the current D3RS evaluation. We indicated that a documented B&W. position would be cu efully considered.

The second issue discussed was that of overpressure protection for the reactor coolant system. B&W's R. Brockman described a graph showing two curves of reactor coolant pressure as a function of temperature.

One curve

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JUL % E 1977 represented the Appendix G (to 10 CFR 50) limiting curve for heatup and cooldown operations, calculated assuming that a BSAR-205 plant had been operating for 32 effective full pcwer years. The other curve showed the allowable reactor coolant system pressure limits whien B&W proposed should be acceptable for the BSAR-205.. system during an overpressure event.

- The reactor coolant temperature range of interest, as presented by B&W, i

was between 305 and 345 degrees Fahrenheit. At lower temperatures the t

OHRS relief valves are designed to limit pressure to less than the Appendix G heatup and cooldown limits. Above 345 degrees Fahrenheit, the pressure allowed by the Appendix G limit is higher than the pressurizer safety valve setpoint.

In the temperature range between 305 and 345 degrees Fahrenheit, the DHRS relief valves will be isolated from the reactor coolant system ~

and the heatup and cooldown pressure limit will be less than the pressurizer safety valve setpoint. B&W proposed that the overpressurization event in this region has such a low probability that the allowable v'essel stress during the incident should be based on emergency condition allowables (modified by fracture toughness requirements) of the ASME Boiler and Pressure Vessel Code.

The allowable pressure curve proposed by B&W is higher than the pressurizer safety valve setpoint in the 305 to 345 degrees Fahrenheit temperature range, thus their claim that -equipment modifications are not necessary to mitigate the consequences of the potential overpressure event.

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Mazetis of the staff stated that if B&W feels that there is a strong argument that the overpressure event and failure to mitigate that o

combined occurrence probability equal to or less than 10-9 vent have a s

per year, then B&W should document that argument. - Staff members present at the meeting agree that sufficient justification for a revised allowable pressure curve, based on emergency conditions and including other changes to remove conservatism, would be difficult to achieve.

B&W committed to early documentation, perhaps Within a week, of their positior dn the two issues discussed in this meeting.

Thomas H. Cox Light Water Reac'ers Branch No. 3 Division of Project Management Enclosures :

1.

Attendance List 2.

DHRS Isolation Capability CC:

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Babcock & Wilcox Company s.--

ATIN: Mr. James H. Taylor dUI 2 21977 Manager of Licensing Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 cc: Washington Public Power Supply system ATIN: Mr. N. O. Strand Mar' aging Director (Acting)

P. O. Box 968 3000 George Washington Way Richland, Washington 99352 Mr. Robert Borsum Beth'esda' Representative Babcock & Wilcox Nuclear Power Generation Division Suite 5515, 7735 Old Georgetown Road Bethesda, Maryland 20014 B. G. Shultz Project Engineer Stone & Webster Ehgineering Corporation P. O. Box 2325 Boston, Massachusetts 02107 Mr. W. E. Kessler Comnonwealth Associates, Inc.

209 East Washington Jackson, Michigan 49201 Robert J. Kafin, Esq.

115 Maple Street Glen Falls, New York 12801 Mr. B. M. Miller Ohio Edison Company 76 South Main Street Akron, Ohio 44308 t

s' ENCLOSURE 1 -

ATTENDANCE LIST-MEETING OF B&W AND NRC STAFF July I?,1977 NAME ORGANIZATION 0.!'Parr CPM L. Riani

. DSS /ASB R. Fitzpatrick DSS /PSB J. Fair SD/EMSB J. Watt DSS /RSB J. Hamilton B&W J. Happell B&W-R. Brockman B&W L. Cartin B&W D. Newton B&W G. Mazetis DSS /RSB S. Burwell DPM/ LWR #2 S. Newberry DSS /RSS

- j T. Novak DSS /RSB T. Cox -

DPM D. Fischer DSS /ASB J. Burdoin DOR /PSB 1

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DERS Isclation Capability s,,

B&W Position 4

The effects of pipe breaks outside containment are to be evaluated per the guidance specified in SRP 3.6.1 and BTP's APCSB 3-1 and HEB 3-1.

B&W's application of these criteria results in 1.

Classification of the DHRS has a noderate energy system with a postulated pipe failure of a through wall leakage in the piping outside containment.

2.

Two power supplies, offsite and the emergency diesels are available.

3.

A single active conponent failure, as defined in Appendix A to APCSB 3-1, is assumed in the systems used to mitigate the pipe failure. The single active failure is pertinent to the ef fected DER train only (Section B.3.6 (3) of BIP APCSB 3-1).

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4.

No assumption of a passive failure other than the piping failure itself.

Mitigation of the postulated piping failure is accomplished from the control room by

1) initiating HPI to make up for RCS leakage and 2) closure of one of the two redundant isolation valves in the DER piping inside containment. 'Since both isolation valves are powered from the same electrical bus, the following interface criteria are proposed to accommodate a single active component failure:

1.

The applicant shall assure that the failure of isolation valves or their ceu124 AM

• electrical pewer supplies are independent of effects of a pipe failure in g

the same fluid system outside containment.

2.

The applicant shall demonstrate that no one single active component failure, assuming both offsite and emergency diesel power are available, shall result in loss of all electrical power to each set of redundant series DHRS isolatic-

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valves while the RCS is operating on the DHRS.

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N-B&W does not consider failure of the electrical bus to be an active failure. This component has no moving parts, and the loss of struccural integrity is precluded f

by definition of a single active component failure given in Appendix A to APCSB 3-1.

1 The postulation of a passive failure, such as the existence of a foreign object creating a 3-phase short, would be an event independent of the postulated piping failure and not a required assumption per the SRP and BTP noted above. Furthermore, n

the probability of such an event is quite remote. Based on the failure data in Appendix III to WASH-1400, the probability of the pipe failure alone is esticated to be N1.75x10~ /yr (Table III-2 of WASH-1400). This event combined with a passive bus

-10 failure further decreases the probability to S3x10

/yr. It is our position that no design changes are warranted to acco=modate such an unlikely event.

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