ML20209H852

Forwards 861021 Safety Evaluation Supporting BWR Owners Group Topical Response to ATWS Rule 10CFR50.62.Rept May Be Ref by Licensees Participating in Owners Group Topical Rept. Plant-specific Inputs Requested by 870331

ML20209H852
Person / Time
Site:	Quad Cities
Issue date:	01/27/1987
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:	Farrar D COMMONWEALTH EDISON CO.
References
TAC-59132, TAC-59133, NUDOCS 8702060153
Download: ML20209H852 (3)

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UNITED STATES NUCLEAR REGULATORY COMMISSION

$ E WASHINGTON, D. C. 20555 o,

• • o ., , , , , +#p January 27, 1987 Docket Nos. 50-254/265 Mr. Dennis L. Farrar Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690

SUBJECT:

PLANT-SPECIFIC ATWS REVIEW GUIDELINES AND IMPLEMENTATION SCHEDULE (MPA A-20 AND TACS 59132,59133)

Re: Quad Cities Nuclear Power Station, Units 1 and 2.

Enclosed is the staff's October 21, 1986, safety evaluation (SE) in support of the BWR Owners Group's licensing topical response to the ATWS rule, 10 CFR 50.62. In accordance with paragraph (c)(6) of 10 CFR 50.67, you are

required to submit sufficient information to demonstrate to the Commission the adequacy of an alternate rod injection (ARI) system, a standby liquid centrol (SLC) system and a reliable reactor coolant recirculating pump trip (RPT) system.

Those licensees who participated in the BWR Owners Group topical report may

reference the topical report in support of their plant-specific submittal.

i For the SLC system, you should choose one of three alternatives and follow the condition stated in the staff's SE. For the ARI system, we suggest you use the checklist found in Appendix A of the staff's SE to address the design details and its conformance with the ARI design objectives and design basis requirements. For the RPT system, you should indicate whether your installed system is the same as the Monticello design or the modified i Hatch design. The staff requires those utilities with the RPT designs to

! submit their schedule to upgrade their designs to either of the above l approved designs or to demonstrate that their present design can perform its function in a reliable manner equivalent to the two approved designs.

g2060153870127 p ADOCK 05000254 PDR

Mr. Dennis L. Farrar -?- January 27, 1987 Technical Specifications should be proposed for ATWS-related components in your plant-specific submittal. In order to support timely closure of this issue we request plant-specific inputs by March 31, 1987. Please advise Tom Rotella, your NRC Project Manager, of any problems or questions.

This request for information was approved by OMB under clearance number 3150-0011.

Sincerely, Original signed by John A. Zwolinski, Director BWR Project Directorate #1 Division of BWR Licensing Enciosure:

Safety Evaluation (AN0:8611030127) cc w/

Enclosure:

See next page DISTRIBUTION: .w/ enclosure DISTRIBUTION: w/o enclosure j  ; Docket File. RBernero NRC PDR EMarinos Local PDR TCollins t

BWD1 Reading HLi TRotella Edordan CJamerson BGrimes Quad Cities File JPartlow JZwolinski OGC-BETH (Info)

NThompson ACRS (10) l l

D53L:BK'D1 DBL:BWD1 CJamerson DBL:BWD1 [ JZwolinski TRotella:ac 01/p-/87 01/rd/87 01/2'1/87 t

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Mr. Dennis L. Farrar Quad Cities Nuclear Power Station Commonwealth Edison Company Units I and 2 cc:

Mr. B. C. O'Brien President Iowa-Illinois Gas and Electric Company 206 East Second Avenue Davenport, Iowa 52801 Mr. Michael I. Miller Isham, Lincoln & Beale Three First National Plaza Suite 5200 Chicago, Illinois 60602 Mr. Nick Kalivianakis Plant Superintendent Quad Cities Nuclear Power Station 22710 - 206th Avenue - North Cordova, Illinois 61242 Resident Inspector U. S. Nuclear Regulatory Commission 22712 206th Avenue North Cordova, Illinois 61242 Chairman Rock Island County Board of Supervisors Rock Island County Court House Rock Island, Illinois 61201 Mr. Michael E. Parker Division of Engineering Illinois Department of Nuclear Safety 1035 Outer Park Drive, 5th Floor Springfield, Illinois 62704 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 799 Roosevelt Road

, Glen Ellyn, Illinois 60137 I

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ENCLOSURE u

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.I NUCLEAR REGULATORY COMMISSION

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l Mr. Terry A. Pickens, Chairman BWR Owners' Group c/o Northern States Power Cor.pany

' 414 Nicollet Hall l

Minneapolis, MN 55401

SUBJECT:

ACCEPTANCE FOR REFERENCING OF LICENSING TOPICAL REPORT NEDE-31096-P, " ANTICIPATED TRANSIENTS WITHOUT SCRAM; RESPONSE TO NRC ATW5 RULE, 10 CFR 50.62" We have completed our review of the subject topical report submitted by

' your letter EWROG-8502 dated January 14, 1986. We find the report to,be acceptable for referencing in license applications to the extent ..

speified and under the limitations delineated in the report and the

/ associated NRC evaluation, which is enclosed. The evaluation defines *he

\ basis for acceptance of the report. .

We do not intend to repeat our review of the matters described in the report and found acceptable when the report appears as a reference in license applications, except to assure that the material presented is applicable to the specific plant involved. Our acceptance applies only to the satters described in the report.

l .

In accordance with procedures established in NUREG-0390, it is requested that the BWR Owners' Group publish accepted versions of this report, proprietary and non proprietary, within three months of receipt of this

' letter. The accepted versions shall incorporate this letter and the enclosed evaluation between the title page and the abstract. The accepted versions shall include a -A (designating accepted) following the report identification symbol.

l Should our criteria or regulations change such that our conclusions as to

( the acceptability of the report are invalidated, the BWR Owners' Group and/or the licensees referencing the topical report will be expected to revise 2

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( Accdssions No. -F'000127 ~

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...............-.......,s. . .

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, Mr. Terry A. Pickens .

and resubmit their respective documentation, or submit justification for the continued effective applicability of the topical report without

. revision of their respective documentation.

Sincerely, bw.

us 1.ainas, Assistant Director Division of BWR Licensing

Enclosure:

As stated cc w/ enclosure:

' Mr. J. M. Fulton, BWR Owners' Group Mr. G. G. Sherwood, General Electric Co.

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e SAFETY EVALUATION OF TOPICAL REPORT (NEDE-31096-P)

"ANTICIPA'TED TRANSIENT WITHOUT SCRAM:

RESPONSE TO AWS RULE.10 CFR 50.62"

1. INTRODUCTION In response to 10 CFR 50.62 " Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATVS) Events for Light-Water-Cooled Nuclear Power Plants", General Electric, on behalf of the. BWR Owners' Group, has published a licensing topical report NEDE-31096-P

, ( " Anticipated Transients Without Scram; Response to NRC ATWS Rule 10 CFR 50.62" which details conceptual designs to satisfy the 10 CFR 50.62 requirements for boiling water reactors. '

This topical report provides the methods for detercining equivalency to the 86 gpm,13-weight percent sodium pentaborate solution of the Standby J

, Liquid Control (SLC) system, provides the design objectives and design basis requirments for the Alternate Rod Injection (ARI) system, and describes the three Recirculation Pump Trip (RPT) designs that are i- currently in use in BWRs.

j 2I PURPOSE '

l (l' The purpose of this SER is to evaluate the acceptability of the proposed conceptual designs to meet the requirements of the ATW5 rule.

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3. BACKGROUND AND CRITERIA -

On July 26, 1984, the Code of Federal Regulations (CFR) was amended to include Section 10 CFR 50.62, " Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants" (known as the "ATV5 Rule"). An ATWS is an expected operational transient (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power) which is accompanied by a failure of the reactor trip system (RTS) to shutdown the rea:: tor. The ATWS rule requires specific improvements in the design and operation of commercial nuclear power facilities to reduce the lik'elihood of failure to shutdown the reactor following anticipated transients, and to mitigate the consequences of an ATWS event.

The systems and equipment required by 10 CFR 50.62 do not have' to meet all of the stringent requirements normally artlied to safety-related equipment. However, this equipment is part of the broader class of structures, systems, and components important to safety defined in the introduction to 20 CFR 50 Appendix A, General. Design Criteria (GDC).

b GDC-2 re:Jires that " structures, systems, and components irportant to safety shall be de:igned, fabricated, erected, and tested to quality standards commensurate with the importance of the sa,fe.y functions to be

[ p e rformed. " Generic Letter 85-06

• Quality Assurance Guidance for ATWS J Equipment that is not Safety Related" details the quality assurance that

[ must be applied to this equipment, i

, In general, the equipment to be installed in accordance with the ATWS rule t

is required to be diverse from the existing RTS, and must be testable at

power. This equipment is intended to provide needed diversity (where only 4

minimal diversity currently exists in the RTS) to reduce the potential for common mode failures that could result in an ATWS leading to unacceptable plant conditions. -

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• The criteria used in evaluating this ' topical report include 10 CFR 50.62

" Rule Considerations Regarding Systems and Equipment criteria" publishe in Federal Register Volume 49, No.124 dated June 26, 1984 and Generic Letter 85-06 " Quality Assurance Guidance for ATWS Equipment that is not Safety Related."

Staff evaluation of the SLC, ARI and RPT features are provided in the following se:tions.

4.

EVALUATION OF STANDBY LIQUID CONTR0t SYSTEM EQUIVALENT CONTROL CAPACITY

4.1 INTRODUCTION

The basic requirement for the standby liquid control system (SLCS) is specified in paragraph (c)(4) of 10 CFR 50.62 (ATWS rule) whic' h states, in part:

"Each boiling water reactor aust have a standby liquid control system (SLCS) capacity control with a minimum to 86 flow capacity and boren content equivalent in .

pentaborate solution." gallons per minute of 13-weight percent spdium Clarification of " equivalent control capacity" was provided in Reference 1 as follows:

(1) The ' equivalent in control capacity" wording was chosen to allow i

flexibility in the implementation of the requirement. For exariple, the equivalence can be obtained by increasing flow rate, boron concentration or boron enrichment.

(2) The 86 gallons per minute and 13 weight percent so:!!um pentaborate vere f values used in NEDE-24222 " Assessment of BWR Mitigation of ATWS, Volumes .

A I and II," December 1979, for BWR/5 and BWR/6 plants with a 251-inch i -

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vessel inside diameter. That different values would be equivalent for smaller plants was recognized in NEDE-24222 in the BWR/4 analysis:

"The flow rates given here are normalized from a 251-inch diameter vessel plant to a 218-inch diameter vessel plant, i.e. , the 66 gpm control liquid injection rate in 218 is equivalent to 86 gpm in a 251. This is done to bound the analysis. .. (pp.2-15 [NEDE-24222])."

(3) The important parameter to consider in establishing equivalence is the

[ time to achieve the necessary boron concentration to bring the reactor to het shutdown.

The minimally acceptable system should show an equivalence in shutdown timing to the generic, reactors studied in NEDE-24222.

4.2 DISCUSSION 4.2.1 SLC system s Desion Basis 3

The generic design basis for the SLC system has historically been to provide a soluble boron (BID) concentration to the core coolant in the reactor vessel sufficient to bring the reactor core to a cold shutdown t

condition within about one or two hours.

j: The ATW5 rule adds injection rate requirements that exceed the generic e design basis.

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Changes to flow rate, solution concentration or boron enrichment, to meet ATWS rule, must not invalidate the original SLC system design basis.

4.2.2 _ Equivalency Consideration .

The ATWS rule relates to hot shutdown rather than to cold shutdown.

Consequently, the time required to deliver to the vessel sufficient boron ;

p to achieve hot shutdown is the important parameter for ATWS.

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. The effective rate of boron injection into the core is the produ pumping capacity (flow rate), solution concentration, baron (810 enrichment, and mixing capacity.

Previously conducted mixing tests were o

accepted by the NRC staff and os a result boron mixing is not a fa determining equivalency to the ATWS rule.

The equivalency relationship is satisfied: requirement can be demonstrated if the follow L

86

• _M251 X C E 1 M T3

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where:

Q = expected SLCS flow rate (gpm)

M = mass of wter in the reactor vessel and rectreulatio system at het rated condition (1bs)

C = IO E=B sodium pentaborate solution concentration (weight perce

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isotope enrichment (19.8 % for natural boron), atom percent Values of M251 may vary somewhat depending on., the M251 design, for (e.g BWR/3/4 = 628,300 lbs, H251 for BWR/5 = 614,300 lbs, and M251 for 8WR/6 =

615,100 lbs).

, The variations are small and, therefore, the ratio of M251/M can be taken as equal to one.

4.2.3 ll Two Pump Operation er Increased Sodium Pentab rate Concentration g.

i For two pump operation with natural enrichment the following rela

! must be satisfied for the total flow rate (Q):

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cr, for increased concentration.of sodium pentaborate (C) the equivalency requirement is:

86 M F FEI (weight percent) where previously defined parameters and natural boron are assumed.

4. 2.4 Beren Enrichment For boron enrichment (E) equivalen.cy the following must be satisfied:

E 2 19.8 X fX fl51 X h (atom percent) 5.

EVALUATION OF STANDBY L1 QUID CONTROL SYSTEM (SLCS) ALTERNATIVES

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5.1 INTRODUCTION

5.1.1 Current Standard System Desion The current standard SLC system design consists of two pumps (each at 43 gpm), designed to operate one pump at a time to inject sodiu*. p ntaborate

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solution into the reactor. The SLC system is initiated remote-manually from the control room.

The solution storage tank and the suction line are heated to prevent precipitation (crystallization) of the sodium pentaborate solution. The concentration of the solution may vary within technical specification limits from about 8% to 21% by weight.

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5.2 SYSTEM ALTERNATIVES 5.2.1 Two-Pump SLC System Operation j

one way to satisfy the 86 ppm equivalency requirement is to operate both SLC pumps simultaneously. Because the pressure drop in the pump discharge

' lines will increase significantly for two pump operation due to increased t flow velocity in the common injection line, the pertaining technical aspects must be evaluated for each SLCS case separately. Licensees adopting this option must confirm that the minimum pump discharge pressure for two pump operation is 91 thin the SLC system components capability at

, the required flow capacity.

If present equipment is found unsatisfactory, modification of the SLC system will be necessary.

i Periodic testing of the system is.also required to assure that the system is capable of performing as intended.

5.2.2 Increased Concentration of Sodiam Pentaborate Solution

' A second way to satisfy the 86 gpm equivalency requirement is to increase  ;

the concentration of sodium pentaborate solution. This design alternative

. could use one or twc pump operation and use natural or enriched boron.

A higher concentration of sodium pentaborate solution would require less '

maintained volume of the solution in the storage tank. The required volume of the solution could be reduced in inverse proportion to the increase in solution concentration while maintaining the current boron concentration required to achieve shutdown.

However, increased concentration of the sodium per.taborate would require .

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maintaining the solution at temperatures higher than the current level, "

(e.g. , 25% solution concentration would require 106' F),' to avoid crystallization of the solution. The higher the temperature of the

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' solution, the higher the heat loss. Therefore, to assure dependable SLC system operation with significantly increased SLC system solution concentrations, special design features such as additional insulation (particularly in the pur.p suction and discharge lines) and heat tracing, and continuous mixing and/or recirculation of the solution in the storage tank will be required.

Licensees adopting this option must provide assurance that boron precipitation will not disabl,e the system or impair the ATW5 mitigation capabilities of the system.

5.2.3 1

Enriched Boron Solutien A third way proposed to satisfy the 86 gpm equivalency requirement is to enrich the boron in the sodium pentaborate solution with the isotope B 0 ,

This design alternative could maintain the current one pump operation and could use the existing solution concentration. ~

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The higher enrichment of BIO may require less maintained volume of solution in the storage tank. The volume could be reduced in inverse proportion to the increase of the enrichment and still maintain sufficient potency to achieve shutdown.

• Surveillance anc positive verification by periodic testing will be required to assure that the correct isotopic concentration is maintained.

5.3 _SUHARY OF THREE ALTERNATIVES The staff reviewed the above proposed three alternatives and concluded that all three proposed alternatives are feasible and with the conditions

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' noted, meet the ATWS rule 10 CFR 50.62 as intended. This is acceptable to us, a

6.

EVALUATION OF ALTERNATE ROD INJECTION (ARI) SYSTEM The basic requirement for ARI system is specified in paragraph (t)(3) of 10 CFR 50.62, which states,

Each boiling water reactor sust have an alternate rod injection (ARI) system that is di.erse (from 'the reactor trip syster) from sensor output

! to the actuation device. The ARI system must have redundant scram air header exhaust valves. The ARI must be designed to perform its function in a reliable manner and be independent (from the existing reactor trip system) from sensor output to the final actuation d' evice."

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The topical report NEDE-31096-P section 3 provides the design objectives and design basis requirements for the ARI system.

6.1 -ARI DESIGN OBJECTIVE i

The ARI provides a path to reactor shutdown which is diverse and l:

independent from the reactor trip system. The automatic signal to i . * **

initiate the ARI function comes from high reactor vessel pressure or low i-reactor vessel water level. The setpoint for ARI lnitiation should be

' chosen such that a normal scram should already have been initiated by the above stated parameters. Following any of these initiation '

signals, the scram air header valves will be opened to reduce air pressure in the header allowing individual scram inlet vales and scram discharge valves to open. The control rod drive units then insert the control blades to shut down the reactor.  ;

In the submittals on BWR ATWS attigation analyses, the General Electric Company has determined that if rod injection motion begins '

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within 15 seconds and is cor.pleted within 25 seconds from ARI initiation time, then the plant safety considerations will be set. The plant safety -

considerations include the maximum temperature limit of the pressure suppression pool (PSP), the maximum containment design loads, and the integrity of the coolable core geometry. The staff fint.: that the ARI design objectives to meet the plant safety considerations as stated in the topical report are consistent with the ATWS rule requirements, and therefore, are acceptable.

• 6.2 ARI DESIGN BASIS REQUIREMENT The NRC has published considerations regarding system and equipment criteria for ATWS in Federal Register Volume 49, No.124 on June 26, 1984.

The topical report has addressed methods of corpliance item by11 tem'in accordance with this guidance.

• The staff's evaluation of these designs basis requirements is provided below. ,

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(1) SAFETY-RELATED REQUIREMENTS (IEEE STANDARD-279) i The ATWS Rule does not require the ARI system to be safety grade, but the implementation must be such that the existing prctection system continues to meet all applicable safety related criteria.

The report notes that the ARI is not required to be safety-related, however, its interface with existing safety related systems fill allow all applicable safety-related criteria to continue to be met. With respect to this design basis requirement, the staff has the following comments:

In order to assure that the existing reactor trip system will c:atinue to '

! seet all safety related criteria, qualified isolators should.be used for f.

iV ARI system interfaces with safety related systems. Electrical independence from the RTS aust be provided from sensor output to final l

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I actuation devices (ARI solenoid valves). Particular emphasis should be

. placed on the method (s) used to qualify the isolators for their particular function.

This should include an analysis and tests which will demonstrate that the isolater will function under the maximum conditions. c The plant specific design should include qualified isolation devices and the qualification information listed in Appendix B should be available for staff audit.

With the above stipulation, the staff finds that this design basis is*

acceptable. -

(2)' REDUNDANCY I

i The ATWS Rule requires that the ARI system must have redundant scram air header exhaust valves, but the ARI system itself does not need to be redundant, j

The generic ARI design indicates that there are redundant scram air header exhaust valves. The report states that although the ARI system does

' not need to be redundant in itself, the ARI performs a function redundant to the backup scram system. The ARI syste: has a different design basis and receives a different initiation signal. All vent paths must function to meet the design basis rod insertion time. The staff finds this acceptable, p'

(3) DIVERSITY FROM EXISTING REACTOR TRIP SYSTEM (RTS) s The ATWS Rule requires that the ARI system should be diverse from the i existing reactor trip system.

c The report states that the ARI system will utilize energize-to-function valves instead of deenergize-to-trip valves. DC powered valvet and %gic Instead of AC power shall be utilized.

Existing Recirculation Pump Trip

(. initiation instrumentation will be used where possible. With respect to this design basis requirement, the staff has the following consents:

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Equipment diversity to the extent reasonable and practicable to minimize the potential for common cause failures is required. This should include i

all diverse reactor trip system instrument channel components excluding I sensors, but including all signal conditioning, and components used to vent the scram air header.

! Even though senscr diversity is not necessary,

! preferred designs will use separate sensors to provide the signals for the

• diverse equipment required ty the ATWS Rule.

Use of the same sensor for the existing reacto- trip system and the diverse equipment would result in inter connections between the two systetts that are difficult to analyze and which could increase the potential for comen cause failures affecting both systems.

Since the sensors for the equipment required by the ATWS Rule do not have to be safety related, there should be considerable flexibility for using existing sensors without using ret: tor trip system sensors.

In cases where existing protection system sensors are used to provide signals to the diverse equipment, particular emphasis 'should be placed on the design of the mathed used to isolate the signal from the existing protection system to minimize the potential for adverse electrical interactions. Existing protection system instrument-sensing lines may be used. Sensors and instrument-sensing lines should be selected such that adverse interaction with existing control systems are avoided.

It is the staff's determination that this design basis should be

• supplemented with instrument channel components (except sensors) that are diverse from RTS.

(4)

ELECTRICAL INDEPENDENCE FROM THE EXISTIN3 REACTOR T The ATWS Rule guidance states that the ARI system is required to be electrically independent from the existing RTS from sensor output to the final actuation device at which point non-safety related circuits must be isolated from safety related circuits. '

The report recommends that the ARI logic be initiated'by sensors (ftcr.

ICCS) that are separate from the RTS. H:vever, the report states that it

'f is acceptable to utilize RTS sensors to initiate the ARI, provided proper s

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isolation is maintained.

- The ARI control power should also be separated from the RTS control power.

The staff concludes that this design basis is acceptable. However, the plant s'pecific design should use qualified isolation devices, and the qualification information identified in Appendix B should be available for staff audit.

(5) H P' YSICAL SEPARATION FROM THE EXISTING REACTOR TRI The ATW5 Rule guidance states that the implementation of the ARI system i

aust be such that separation criteria applied to the existing protection

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system are not violated.

i The report states that the safety-related RTS is separated from the non-1E

[ ARI, as required by IEEE-279. The ARI, even if it has been installed as a IE system, is physically separated from the RTS. The staff finds this acceptable. ,

1 i

(6) ENVIRONMENTAL QUALIFICATION

• The ATWS Rule guidance states that the qualification of the ARI system is for anticipated operational occurrences only, not fo'r accidents.

The report states that qualification of new equipment will be to I

temperature, pressure, humi:ity and radiation levels associated with

' Anticipated Operational Occurrences, not Design Basis Accidents (LOCA &

HELB).

Equipment must be qualified to conditions during an ATWS event up to the time that the ARI function is completed. The staff finds this

! acceptable.

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(7) SEISMIC QUALIFICATION

• No seismic qualification is required for the ARI system harhare.

(8) QUALITY ASSURANCE hRC Generic Letter 85-06 dated April 16, 1985 provides quality assurance guidance for the ARI system.

• The report states that the NRC Generic Letter 85-06,

• Quality Assurance

! Guidance for ATW5 Equipment that ia not Safety Related" will be complied with on a plant-specific basis.  ;

(9) SAFETY RELATED (IE) POWER SUPPLY p

The ATWS Rule guidance states that the ARI system aust be capable of performing its safety functions wit . loss of offsite power, and that the power source should be independent from existing reactor trip system.

The report statas that the ARI system controls, instrumentation and i

solenoid valves are powered from non-divisional, non-interruptible DC

• power independent from RTS power. This power source allows the ARI to perform its function during any loss-of-offsite power event. Diesel generators are not sufficient to serve as a backup power source to meet the ARI function for the loss-of-offsite power event because of the l time delay beforat power is available. Safety-related power supplies, j

separate from the RTS power source and available during loss-of-offsite power are acceptable for powering ARI if the ARI system is IE or non-1E.

as long as the ARI system is properly isolated from the safety-related power supply.' The staff endorses the above design basis with the following ,

additional guidance:

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t It is the staff's determination that a preferred design will have an ARI l

• power source totally separated from the RTS power source. If the ARI I

system has to use a safety related power supply through " proper 1 L isolation," then two qnlified Class IE breakers" in series with proper i relay coordination should be provided for the isolation function., ("Two

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fuses in series or a combination of one fuse and one breaker will also be acceptable.)

(10) TESTABILITY AT POWER The ATWS Rule guidance states that the ARI system should be testable at

! power.

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In response to the staff's request for additional information, the BWR Owners' Grcup clarified the generic ARI design as follows: Th'e generic ARI system has been designed to permit maintenance repair, test or

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calibration of the system logic and instrumentation up to but not 1

including the final trip devices. This was achieved in one of two ways.

(1) Use of a redundant 2 out-of-2 logic arrangement. Each individual level and pressure instrument can be tested during plant operation without initiating the ARI system since two level or two pressure ,

signals must be present in one channel to. complete the signal.

L i L' (2) Use of a parallel air header supply block valve and vent valve arrangement. One block valve and one vent valve are controlled i

thrcugh one channel and the second block valve and vent valve through a second channel. Two high pressure or two low water level signals must be present in a channel to close one block valve and to open one vent valve. Since both block valves are required to close and both vent valves are required to open for the ARI fur.: tion, each channel  ;

can be tested independently without resulting in an inadvertent ARI initiation. '

The staff finds either arrangement acceptable.

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( n ) INADVERTENT ACTUATION -

The ATWS Rule guidance states that the inadvertent ARI actuation which challenges other safety systems should be minia12ed.

The report states that the setpoints chosen for actuation of this function shall be such that normal reactor scram from the RTS would be expected to occur prior to or concurrent with the ARI function. In response to the staff's request for additional information, the BWR Dwners' Group clarified that redundant coincident logic will be used for ARI initiation ic;;ic. The staff finds this accep' table.

(12) ADDITIONAL FEATURES IN THE GENERIC DESIGN E

In response to the staff's request for additional information'(Reference l

' 2), the BWR Dwners' Group provided additional inforzation regarcing the g

'eneric desigr. in Reference 3.

(a) Manual Initiation The generic ARI system has the capability of being manually initiated from

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• a location in the control room. The staff finds this acceptable.

(b) Information Readout In the design of the ARI system for each BWR plant, consideration has been given to providing adequate readout information for the control room operator.

Due to differences in control room layout, signal availability, and alternate information for inferring status, no single generic design of control room information has been specified. One operator infomation  ;

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provision in common is that each plant provides a means for informing the operator that an ARI has been initiated.

The staff finds this acceptable. The plant specific design should include the ARI system information readout features and this information should be available for staff audit.

c) Completion of Protective Action Once it is Initfa'ted f

The generic ARI system has been designed so that once it is initiated, the protective action will go to completion. Either the automatic or manual actuation signals in the generic ARI system design " seal-in" to assure that all control rods have time to fully insert.

The reset of the ARI function will be prohibited for the duratic'n of the i ,( seal-in time either automatically or by administrative contrels. The ARI function can be manually reset by the operator after completion of the seal-in time if the automatic signals have cleared. No automatic return 3

to normal operation is provided. The staff finds this acceptable.

I (d) Maintenance Bypass and the Means for Bypassing i

L The generic ARI system has been designed to permit maintenance repairs, test or calibration of the system logic and instrumentation up to but not including the final trip d cites. Maintenance bypasses are not requirev.

The staff finds this acceptable. However, it is the staff's determination that if a design needs a maintenance bypass, the means for bypassing should be accomplished with a permanently installed bypass switch or similar device. The use of a maintenance bypass should not involve lifting leads, pulling fuses, or tripping breakers or physically blocking  ;

( relays. The bypass status must be automatically and continuously indicated in the main control room. ..

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6.3, CONCLUSIONS ON AR1 SYSTEM ~

Based on its review, the staff concludes that the ARI design basis requirements stated in the topical report NEDD 330g6-P in conjunction with the staff requirements identified above are in general compliance with ATWS Rule 10 CFR 50.62 paragraph (c)(3) and the guidance regarding system and equipment specifications published in Federal Register Volume 4g, No. 124 dated June 26, 1g84.

To facilitate prompt review of plant specific ARI designs, the staff has developed Appendix A to this SER which itemizes the AR1 features approved by the staff. The licensees or applicants who commit to fully implement or have implemented an ARI design incorporating these features covered in the Appendix A will be considered to be in conformance with the ATW5 Rule 10 CFR E0.62 paragraph (c)(3) on ARI requirements.

The staff will review .pecific plant design ARI systems for licensees or applicants who do not fully corply with the specified features covered in Appendix A.

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7. EVALUATION OF REClRCUU* ION PUW.P TRIP (RPT) SYSTEN i

The basic requirement for the RPT system is specified in paragraph (c)(5) of 10 CFR 50.62, which states, "Each boiling water reactor must have equipment to trip the reactor coolant recirculating pumps autor,atically under conditions indicative of an ATWS. This e reliable manner."quipment must be designed to perform its function in a The topical report discusses three standard RPT designs:  ;

(1)ModifiedHatchDesign .

(. (2)MonticelloDesign

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(3) Original BWR/4 Design .

The Modified Hatch Design employs two trip coils in each recirculation system motor generator set generator field breaker. The input logic is from a one-out-of-two low reactor vessel water level signal or a one-out-of-two high reactor pressure signal. The Monticello Design also employs two trip coils in each recirculation systen moto generator set generator field breaker. The input logic is fres a two-out-of-two low reactor vessel water level signal or a two out-of-two high reactor j pressure signal. The third design (Original BWR/4 design) employs a

single trip cell in the M-G Set Drive Motor Feeder Breaker. Pump trip logic A&C trips the "A" pump and pump trip logic S&D trips the 8" pump.

In 1979, the Commission approved both the Monticello design and the

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modified Hatch design for the RPT system. The licensees were asked to 4 j

respond within 90 days with their schedules for implementation of an RPT l.

system of either-the Monticello or the sodified-Hatch design. The staff has net reviewed plant specific RPT designs for RPT systems installed

{ prior to 1979. In 1979, the plant specific RPT design reviews were p.erformed by an NRC contractor (EG&G) for 11 plants which were to install RPT systems retting the NRC criteria. A summary report (Reference 4) was published which indicated 9 out of Il plants utilize the Monticello design, and two plants did not identify their RPT design. A survey was performed by the staff for all the operating BWR plants on RPl status.

The survey indicated that all the operating BWRs have a RPT system installed except Big Rock Point which was granted an exemption.

After the ATWS Rule was published in 1984, the Nebraska Public Power District (Cooper Station) committed to upgrade its present recirculation' I pump trip system to the Monticello design. The Tennessee Valley Authority :

(Browns Ferry Units 1, 2, & 3) alsc indicated in a recent meeting that the RPT design will be upgraded. The staff will require those utilities that have not upgraded the RPT system to either Monticello or modified Match l

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=

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6 design to submit their proposed upgrade plan or to demonstrate that their present design can perform its function in a reliable manner equivalent to the Monticello design or the modified Hatch design.

The topical report table 4-1 note 2 states that some plants use end-of-cycle (EOC) RPT breakers for the ATW5 trip. This arrangement is acceptable provided that qualified isolators are used between the "EOC-RPT" signal and the "ATVS-RPT" signal to maintain electrical independence between the reactor trip system and the ATW5 system.

7.1 CONCLUSION

S ON RPT

• The Monticello design and the modified Hatch design are acceptable for reference. The staff will require these utt11 ties with other designs to submit their proposed upgrade plan or to demonstrate that the'* present

( design can perform its function in a reliable manner equivalent to the two approved designs.

8. TECHNICAL SPECIFICATION i

Technical specification requirements for ATW5 related components must be l

• addressed by plant' specific submittals. .

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APPENDIX A

. CHECKLIST FDR PLANT SPECIFIC REVIEW OF ALTERNATE RDD INJECTIDN SYSTEM (ARI)

Conformance with ARI SER

2. ARI system function time Rod injection action will begin within 15 seconds and be completed within 25 seconds from ARI initiation

2. Safety-related requirements (a) Class IE isolators are used to interface with safety-related systems (b) Class IE isolators are powered from a '

Class IE scarce (c) Isolator qualification documents are available for staff audit ,

i 3. Redundancy The ARI system performs a function redundant to the backup scram system

4. Diversity from existing RTS l

(a) ARI system is energize-to-function (b) ARI system uses DC powered valves (c) Instrument channel components (excluding sensors but including all signal conditioning and isolation devices) are diverse from the the existing RTS components.

5. Electrical independence from the existing RTS (a) ARI actuation logic separate from RTS logic (b) ARI circuits are isolated from safety related circuits

• ._.,., .c m-w--my~ ~ ~ - - -- - - - - ' ' ' " ^ ' ' " " "

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6. .

Physical separation from the existing RTS

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(a) ARI system is physically separated from RTS 1

[. Environmental Qualification

.. - ARI equipments are qualified to conditions during an A1WS event up to the time the ARI function is completed

8. - Quality Assurance i

(a) Comply with Generic Letter 85-06 .

-5 9. Safety related power supply r (a) ARI system power independent from RTS

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-(b) ARI system can perfom'its function during any loss-of-offsite power event s 20. Testability'st P'dwer

! '. (a) ARI testable at power r

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(b) Bypass features conform to bypass i

., criteria used in RTS

22. Inadvertent Actuation .

. (a) ARI Actuation setpoints will not challenge scram i

(b) Coincident logic is utilized in ARI design

22. Manue' Initiation

}- (a) Manual initiation capability is provided '

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13. Information Readout (a) Information readeut is provided in main control room

24. Completion of protective action once it is initiated O

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