ML20207L931

Requests Submittal of Info Demonstrating Adequacy of Alternate Rod Insertion,Standby Liquid Control & Recirculation Pump Trip Sys.Deviations from BWR Owners Group Alternatives Must Be Justified

ML20207L931
Person / Time
Site:	Hatch, San Onofre
Issue date:	01/08/1987
From:	Rivenbark G Office of Nuclear Reactor Regulation
To:	James O'Reilly GEORGIA POWER CO.
References
NUDOCS 8701130039
Download: ML20207L931 (3)

Text

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+ Jcnuary 8, 1987 Dockets Nos. 50-321/361 Mr. James P. O'Reilly Senior Vice President -

Nuclear Operations Georgia Power Company Post Office Box 4545 Atlanta, Georgia 30302

SUBJECT:

ATWS RULE (10 CFR 50.62): PLANT SPECIFIC REVIEWS Re: Edwin I. Hatch Nuclear Plant, Units 1 and 2 Paragraph (C)(6) of 10 CFR 50.62 requires that all BWR licensees submit information sufficient to demonstrate the adequacy of the systems addressed in Paragraphs (C)(3), (C)(4), and (C)(5), for their respective facilities. These systems include the alternate rod insertion (ARI), standby liquid control (SLC), and recirculation pump trip (RPT) systems.

Since your utility was a participant in the BWR Owners Group licensing topical report (NEDE-31096-P) submitted January 14, 1986, in response to 10 CFR 50.62, you may reference this document to support your plant specific submittal. For your information, we have enclosed a copy of the staff's Safety Evaluation (SE) of the topical report (Enclosure 1).

With regard to the SLC system, your submittal should address the conditions stated in the staff's SE for the system alternative you have selected. For the ARI system, you may use the checklist contained in Appendix A of the SE to address the details of your design and its conformance to ARI design basis requirements and objectives. Regarding the RPT system, you should indicate whether the system installed at Hatch Units 1 and 2 is based on the "Monticello design" or the " modified Hatch design". If your system is not based on either of these approved designs, you should submit your schedule for upgrading the system to one of these designs, or else demonstrate that the existing system can perform its function with equivalent reliability.

If your utility has chosen to deviate from the approved design alternatives contained in the BWR Owners Group topical report, you are requested to submit detailed design information. This information should address all aspects of 10 CFR 50.62 " Rule Considerations Regarding Systems and Equipment Criteria" published in Federal Register Volume 49, No. 124 dated June 26, 1984, and Generic Letter 65-06, " Quality Assurance Guidance for ATWS Equipment that is not Safety Related".

8701130039 87olos PDR P

ADOCK 05000321 PDR

Your plant specific submittal, in addition, should include any proposed Technical Specifications associated with the ATWS-related components.

In order to complete our review in a timely manner with respect to the implementation schedule for the ATWS rule, we request that you provide your response for Unit 1 no later than 45 days from the date of receipt of this letter and no later than 90 days from the date of receipt of this letter for Unit 2.

This request for information was approved by 0MB under clearance number 3150-0011.

Sincerely,

%31 sip ~$ q George Rivenbark, Project Manager BWR Project Directorate #2 Division of BWR Licensing

Enclosure:

As stated cc w/ enclosure See next page DISTRIBUTION w/ enclosure DecketcF.Hes a NRC PDR Local PDR GRivenbark SNorris Plant Files w/o enclosure RBernero EJordan BGrimes JPartlow NThompson ACRS(10)

OGC-Bethesda /N f f il l DB : D#2 DBL { 80#2 DBL #2 #2 S rs GRhynbark Di dr Ise 1/f(/87 1/ y /87 13 /87 1/1/87 0FFICIAL RECORD COPY

Mr. J. P. O'Reilly Edwin I. Hatch Nuclear Plant, Georgia Power Company Units Nos. I and 2 cc:

Bruce W. Chruchill, Esquire Shaw, Pittman, Potts & Trowbridge 2300 N Street, N.W.

Washington, D.C. 20037 Mr. L. T. Gucwa Engineering Department Georgia Pwer Company Post Office Box 4545 Atlanta, Georgia 30302 Mr. H. C. Nix, Jr. , General Manager Edwin I. Hatch Nuclear Plant Georgia Power Company Post Office Box 442 Baxley, Georgia 31513 Mr. Louis B. Long Southern Company Services, Inc. .

Post Office Box 2625 Birmingham, Alabama 35202 Resident Inspector U.S. Nuclear Regulatory Commission Route 1, Post Office Box 279 Baxley, Georgia 31513 Regional Administrator, Region II U.S. Nuclear Regulatory Commission, 101 Marietta Street, Suite 2900 Atlanta, Georgia 30303 Mr. Charles H. Badger Office of Planning and Budget ,

Room 610 270 Washington Street, S.W.

Atlanta, Georgia 30334 Mr. J. Leonard Ledbetter, Commissioner Department of Natural Resources 270 Washington Street, N.W.

Atlanta, Georgia 30334 Chairman Appling County Commissioners County Courthouse Baxley, Georgia 31513

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n i e  % UNITED STATES I  % NUCLEAR REGULATORY COMMISSION y .j m senwarow.o.c. noses g,.....j October 21,1986 l

T Mr. Terry A. Pickens, Chairman SWR Owners' Group c/o Northern States Power Company 414 Nicollet Mall

. Minneapolis, MN 55401

Dear Mr. Pickens:

i

SUBJECT:

ACCEPTANCE FOR REFERENCING 0F LICENSING TOPICAL REPORT NEDE-31096-P, " ANTICIPATED TRANSIENTS WITHOUT SCRAM; RESPONSE TO NRC ATWS RULE, 10 CFR 50.62" We have completed our review of the subject topical report submitted by l your letter 8WROG-8602 dated January 14, 1986. We find the report.to be acceptable for referencing in license applications to the extent i specified and under the limitations delineated in the report and the associated NRC evaluation, which is enclosed. The evaluation defines the j 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 i

applicable to the specific plant involved. Our acceptance applies only to the matters described in the report.

. In accordance with procedures established in NUREG-0390, it is requested

that the BWR Owners' Group publish accepted versions of this report, i

proprietary and non proprietary, within three months of receipt of this I

letter. The accepted versions shall incorporate this letter and the 1 enclosed evaluation between the title page and the abstract. The accepted versions shall include a -A (designating accepted) following j the report identification symbol.

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

, Mr. Terry A. Pickens -2+

and resub it their respective documentation, or submit justification for the continued effective applicability of the topical report without revision of their respective documentation.

Sincerely, s

$4 W us Lainas, Assistant Director Division of BWR Licensing

Enclosure:

As stated j

cc w/ enclosure:

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

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+f'purgIo 'g UNITED STATES

! o NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20655

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

" ANTICIPATED TRANSIENT WITHOUT SCRAM:

RESPONSE TO ATWS RULE, 10 CFR 50.62" 1

1. INTRODUCTION In response to 10 CFR 50.62 " Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) 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 determining equivalency to the 86 gpm, 13-weight percent sodium pentaborate solution of the Standby 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 currently in use in BWRs.

2. PURPOSE ThepurposeofthisSERistoevaluatethe,$cceptabilityoftheproposed conceptual designs to meet the requirements of the ATWS 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 "ATWS 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 reactor. The ATWS rule requires specific improvements in the design and operation of commercial nuclear power facilities to reduce the likelihood 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 applied 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 10 CFR 50, Appendix A, General Design Criteria (GDC).

GDC-1 requires that " structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed." Generic Letter 85-06 " Quality Assurance Guidance for ATWS Equipment that is not Safety Related" details the quality assurance that must be applied to this equipment.

In general, the equipment to be installed in accordance with the ATWS rule 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 j 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" published 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 sections.

4. EVALUATION OF STANDBY LIQUID CONTROL 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) which states, in part:

"Each boiling water reactor must have a standby liquid control system (SLCS) with a minimum flow capacity and boron content equivalent in control capacity to 86 gallons per minute of 13-weight percent sodium pentaborate solution."

Clarification of " equivalent control capacity" was provided in Reference 1 as follows:

(1) The " equivalent in control capacity" wording was chosen to allow flexibility in the implementation of the requirement. For example, the equivalence can be obtained by increasing flow rate, boron concentration or boron enrichment.

(2) The 86 gallons per minute and 13-weight percent sodium pentaborate were values used in NEDE-24222, " Assessment of BWR Mitigation of ATWS, Volun.::

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

4 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 hot 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 Design Basis The generic design basis for the SLC system has historically been to provide a soluble boron (B10) concentration to the core coolant in the reactor vessel sufficient to bring the reactor core to a cold shutdown condition within about one or two hours.

The ATWS rule adds injection rate' requirements that exceed the generic design basis.

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 to achieve hot shutdown is the important parameter for ATWS.

e The effective rate of boron injection into the core is the product of pumping capacity (flow rate), solution concentration, boron (B10) enrichment, and mixing capacity. Previously conducted mixing tests were accepted by the NRC staff and as a result boron mixing is not a factor in determining equivalency to the ATWS rule.

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

C E 1 R-86 X M251M X 13 X 19.8 g

4- where: Q = expected SLCS flow rate (gpm)

M = mass of water in the reactor vessel and recirculation system at hot rated condition (lbs)

C = sodium pentaborate solution concentration (weight percent)

E=B 10 isotope enrichment (19.8 % for natural boron), atom percent Values of M251 may vary somewhat depending on the design, (e.g., M251 for BWR/3/4 = 628,300 lbs, M251 for BWR/5 = 614,300 lbs, and M251 for BWR/6 =

615,100 lbs). The variations are small and, therefore, the ratio of M251/M can be taken as equal to one, i

4.2.3 Two Pump Operation or Increased Sodium Pentaborate Concentration For two pump operation with natural enrichment the following relationship must be satisfied for the total flow rate (Q):

M 13 Q 2 86 X M251 X

7 (gp,)

._ - - - ~ _ . _ _ _ - - - ,_. - . _ _ . - - ~ _

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or, for increaseu concentration of sodium pentaborate (C) the equivalency requirement is:

251 (weight percent) where previously defined parameters and natural boron are assumed.

4.2.4 Boron Enrichment For boron enrichment (E) equivalency the following must be satisfied:

86 M 13 A lO X X Q- H251 C (atom percent)

5. EVALUATION OF STANDBY LIQUID CONTROL SYSTEM (SLCS) ALTERNATIVES

5.1 INTRODUCTION

5.1.1 Current Standard System Desian The current standard SLC system design consists of two ptmps (each at 43 gpm), designed to operate one pump at a time to inject sodium pentaborate 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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e 5.2 SYSTEM ALTERNATIVES 5.2.1 Two-Pump SLC System Operation One way to satisfy the 86 gpm 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 flow velocity in the common injection line, the pertai..'.ng 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 within the SLC system components capability at the required flow capacity.

If present equipment is found unsatisfactory, modification of the SLC system will be necessary. 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 Sodium 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 two 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 baron concentration required to achieve shutdown.

4 However, increased concentration of the sodium pentaborate would require maintaining the solution at temperatures higher than the current level, (e.g., 26% solution concentration would require 106 F), to avoid crystallization of the solution. The higher the temperature of the

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 i

(particularly in the pump 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 disable the system or impair the ATWS mitigation capabilities of the system.

5.2.3 Enriched Boron Solution 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 10 ,

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

The higher enrichment of B 10 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 and positive verification by periodic testing will be required to assure that the correct isotopic concentration is maintained.

5.3

SUMMARY

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

o notad, meet the ATWS rule 10 CFR 50.62 as intended. This is acceptable to us.

6. EVALUATION OF ALTERNATE ROD INJECTION (ARI) SYSTEM The basic requirement for ARI system is specified in paragraph (C)(3) of 10 CFR 50.62, which states, "Each boiling water reactor must have an alternate rod injection (ARI) system that is diverse (from the reactor trip system) 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 device."

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 The ARI provides a path to reactor shutdown which is diverse and independent from the reactor trip system. The automatic signal to initiate the ARI function comes from high reactor vessel pressure or low reactor vessel water level. The setpoint for ARI initiation 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 mitigation analyses, the General Electric Company has determined that if rod injection motion begins

1

_ 10 within 15 seconds and-is completed within 25 seconds from ARI~ initiation time, then the plant safety considerations will be met. 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 finds 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 compliance item by item in accordance with this guidance.

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

(1) SAFETY-RELATED REQUIREMENTS (IEEE STANDARD-279)

The ATWS Rule does not require the ARI system to be safety grade, but the implementation must be such that the existing protection 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 will 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 continue to meet all safety related criteria, qualified isolators should be used for

. ARI system interfaces with safety related systems. tiectrical

! independence from the RTS must be provided from sensor output to final

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 isolator will function under the maximum case fault conditions. 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 The ATW5 Rule requires that the ARI system must have redundant scram air l

header exhaust valves, but the ARI system itself does not need to be redundant.

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 system has a different design basis l and receives a different initiation signal. All vent paths must function to meet the design basis rod insertion time. The staff finds this acceptable.

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

The report states that the ARI system will utilize energize-to-function l valves instead of deenergize-to-trip valves. DC powered valves and logic

{ 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 comments:

t

Equipment diversity to the extent reasonable and practicable to minimize the potential for common cause failures is required. This should include all diverse reactor trip system instrument channel components excluding sensors, but including all signal conditioning, and components used to vent the scram air header. Even though sensor diversity is not necessary, preferred designs will use separate sensors to provide the signals for the diverse equipment required by the ATWS Rule. Use of the same sensor for the existing reactor trip system and the diverse equipment would result in inter-connections between the two aystems that are difficult to analyze and which could increase the p'tential o for common cause failures affecting both systems. Since the sensors for the equipment required by the ATWS Rale do not have to be safety related, there should be considerable flexibility for using existing sensors without using reactor trip system sensors. In cases where existing protection ::ystem sensors are used to provide signals to the diverse equipment, particular emphasis should be placed on the design of the method 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 EXISTING REACTOR TRIP SYSTEM (RTS)

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 (from ECCS) that are separate from the RTS. However, the report states that it is acceptable to utilize RTS sensors to initiate the ARI, provided proper

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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 specific design should use qualified isolation devices, and the qualification information identified in Appendix B should be available for staff audit.

(5) PHYSICAL SEPARATION FROM THE EXISTING REACTOR TRIP SYSTEM The ATWS Rule guidance states that the implementation of the ARI system must be such that separation criteria applied to the existing protection system are not violated.

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 1E system, is physically separated from the RTS. The staff finds this acceptable.

(6) ENVIRONMENTAL QUALIFICATION The ATWS Rule guidance states that the qualification of the ARI system is for anticipated operational occurrences only, not for accidents.

The report states that qualification of new equipment will be to temperature, pressure, humidity 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.

O (7) SEISMIC QUALIFICATION No seismic qualification is required for the ARI system hardware.

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

The report states that the NRC Generic Letter 85-06, " Quality Assurance Guidance for ATWS Equipment that is not Safety Related" will be complied with on a plant-specific basis.

(9) SAFETY RELATED (IE) POWER SUPPLY The ATWS Rule guidance states that the ARI system must be capable of performing its safety functions with loss of offsite power, and that the power source should be independent from existing reactor trip system.,

The report states that the ARI system controls, instrumentation and 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 time delay before power is available. Safety-related power supplies, separate from the RTS power source and available during loss-of-offsite power are acceptable for powering ARI if the ARI system is 1E 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:

0 It is the staff's determination that a preferred design will have an ARI power source totally separated from the RTS power source. If the ARI system has to use a safety related power supply through " proper isolation," then two qualified Class 1E breakers

• in series with proper relay. coordination should be provided for the-isolation function. (*Two 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.

In response to the staff's request for additional information, the BWR Owners' Group clarified the generic ARI design as follows: The generic ARI system has been designed to permit maintenance repair, test or calibration of the system logic and instrumentation up to but not 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.

(2) Use of a parallel air header supply block valve and vent valve arrangement. One block valve and one vent valve are controlled through one channel and the s.econd 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 function, each channel can be tested independently without resulting in an inadvertent ARI initiation.

The staff finds either arrangement acceptable.

O (11) INADVERTENT ACTUATION The ATWS Rule guidance states that the inadvertent ARI actuation which challenges other safety systems should be minimized.

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 Owners' Group clarified that redundant coincident logic will be used for ARI initiation logic. The staff finds this acceptable.

(12) ADDITIONAL FEATURES IN THE GENERIC DESIGN In response to the staff's request for additional information (Reference 2), the BWR Owners' Group provided additional information regarding the generic design in Reference 3.

(a) Manual Initiation The generic ARI system has the capability of being manually initiated from a location in the control room. The staff finds this acceptable.

(b) Information Res'out In the design of the ARI system for each BWR plant, consideration has been given to providing adequate readout informatio.- 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 information

• I 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 Initiated 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 duration of the seal-in time either automatically or by administrative controls. 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 to normal operation is provided. The staff finds this acceptable.

(d) Maintenance Bypass and the Means for Bypassing 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 devices. Maintenance bypasses are not required.

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.

6.3 CONCLUSION

S ON ARI SYSTEM Based on its review, the staff concludes that the ARI design basis requirements stated in the topical report NE00-31096-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 49, No. 124 dated June 26, 1984.

To facilitate prompt review of' plant specific ARI designs, the staff has developed Appendix A to this SER which itemizes the ARI 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 ATWS Rule 10 CFR 50.62 paragraph (c)(3) on ARI requirements.

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

7. EVALUSTION OF RECIRCULATION PUMP TRIP (RPT) SYSTEM 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 automatically under conditions indicative of an ATWS. This equipment must be designed to perform its function in a reliable manner."

The topical report discusses three standard RPT designs:

(1) Modified Hatch Design (2) Monticello Design

(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 system motor generator set generator field breaker. The input logic is from a two-out-of-two low reactor vessel water level signal or a two-out-of-two high reactor pressure signal. The third design (Original BWR/4 design) employs a single trip coil in the M-G Set Drive Motor Feeder Breaker. Pump trip logic A&C trips the "A" pump and rump trip logic B&D trips the "B" pump.

In 1979, the Commission approved both the Monticello design and the modified Hatch' design for the RPT system. The licensees were asked to respond within 90 days with their schedules for implementation of an RPT system of either the Monticello or the modified-Hatch design. The staff has not reviewed plant specific RPT designs for RPT systems installed prior to 1979. In 1979, the pignt specific RPT design reviews were performed by an NRC contractor (EG&G) for 11 plants which were to install RPT systems meeting the NRC criteria. A summary report (Reference 4) was published which indicated 9 out of 11 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 RPT 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 pump trip system to the Monticello design. The Tennessee Valley Authority (Browns Ferry Units 1, 2, & 3) also 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 Hatch

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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 (E0C) RPT breakers for the ATWS trip. This errangement is acceptable provided that qualified isolators are used between the "EOC-RPT" signal and the "ATWS-RPT" signal to maintain electrical independence between the reactor trip system and the ATWS system.

7.1 CONCLUSION

S ON RPT Tne Monticello design and the modified Hatch design are acceptable for rsference. The staff will require those utilities with other designs to submit their proposed upgrade plan or to demonstrate that their present design can perform its function in a reliable manner equivalent to the two approved designs.

8. TECHNICAL SPECIFICATION Technical specification requirements for ATWS related components must be addressed by plant specific submittals.

I APPENDIX A CHECKLIST FOR PLANT SPECIFIC REVIEW OF ALTERNATE R0D INJECTION SYSTEM (ARI)

Conformance with ARI SER

- 1. ARI system function time Rod injection motion 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 source (c)Isolatorcualificationdocumentsare

available for staff audit

3. Redundancy The ARI system performs a function redundant to the backup scram system 4

4. Diversity from existing RTS (a)ARIsystemisenergize-to-function (b) ARI system uses DC powered valves (c)Instrumentchannelcomponents(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) AR1 circuits are isolated from safety related circuits

6. Physical separation from the existing RTS (a) ARI system is physically separated from RTS

7. Environmental Qualification ARI equipments are qualified to conditions during an ATWS event up to the time the ARI function is completed

8. Quality Assurance (a) Comply with Generic Letter 85-06

9. Safety-related power supply (a) ARI system power independent from RTS (b) ARI system can perform its function during any loss-of-offsite power event

10. Testability at Power (a) ARI testable at power (b) Bypass features conform tc bypass criteria used in RTS

11. Inadvertent Actuation (a) ARI Actuation setpoints will not challenge scram (b) Coincident logic is utilized in ARI design

12. Manual Initiation (a) Manual initiation capability is provided

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

14. Completion of protective action once it is initiated r

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APPENDIX B QUALIFICATION INFORMATION ON ISOLATION DEVICES

a. For the type of device used to accomplish electrical isolation, describe the specific testing performed to demonstrate that the device is acceptable for its applications). This description should include elementary diagrams when necessary to indicate the test configuration and how the maximum credible faults were applied to the devices,

b. Data to verify that the maximum credible faults applied during the test were the maximum voltage / current to which the device could be exposee, and define how the maximum voltage / current was determined,

c. Data to verify that the maximum credible fault was applied to the output of the device in the transverse mode (between signal and return) and other faults were considered (i.e., open and short circuits).

d. Defir.e the pass / fail acceptance criteria for each type of device.

e. Provide a commitment that the isolation devices comply with the environ-ment qualifications (10 CFR 50.49) and with the seismic qualifications which were the basis for plant licensing,

f. Provide a description of the measures taken to protect the safety systems from electrical interference (i.e., Electrostatic Coupling, EMI, Common Mode and Crosstalk) that may be generated by the ATWS circuits.

g. Provide information to verify that the Class 1E isolator is powered from a Class IE source.

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REFERENCES

1. USNRC Generic Letter 85-03, " Clarification of Equivalent Control Capacity for Standby Liquid Control Systems," January 28, 1985.

2. Letter from M. W. Hodges to T. A. Pickens dated May 29, 1986; Request for Additional Information on Licensing Topical Report NEDE-31096-P,

" Anticipated Transients Without Scram; Response to h4C ATWS Rule 10 CFR 50.62".

3. Letter from T. A. Pickens to M.-W. Hodges deted July 30, 1986; Request for Additional Information on Licensing Topical Report NEDE-31096-P

" Anticipated Transients Without Scram; Response to NRC ATWS Rule 10 CFR 50.62".

4 UCID --18127, " Summary of Boiling Water Reactor Licensee's Commitments to Install Recirculation Pump Trip Systems," July, 1979, by EG&G Energy Measurements Group, San Ramon Operations.

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