ML19337A171

Forwards Request for Addl Info Re RCS Vents,Rhr Sys & Bypass,Override & Reset Circuits of Engineered Safety Features

ML19337A171
Person / Time
Site:	Farley
Issue date:	08/25/1980
From:	Tedesco R Office of Nuclear Reactor Regulation
To:	Clayton F ALABAMA POWER CO.
References
NUDOCS 8009090128
Download: ML19337A171 (18)

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'c, UNITED STATES lV i

NUCLEAR REGULATORY COMMISSION

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i August 25, 1980 Docket No. 50-364 l

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Mr. F. L. Clayton, Jr., Senior Vice President Alabama Power Company Post Office Box 2641 Birmingham, Alabama 35291

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION FOR FARLEY 2 OPERATING LICENSE APPLICATION i

As a result of our continuing review of the operating license application for the Joseph M. Farley Nuclear Plant Unit 2, we have developed the enclosed requests for additional infomation.

Please provide the information requested in the enclosures. Our review schedule is based on the assumption that the additional information will be available for our review by September 8,1980.

If you cannot meet this date, please inform us within 7 days after receipt of this letter so that we may revise our scheduling.

Sincerely,

& %g.,,0 Robert L. Tedesco Assistant Director for Licensing Division of Licensing

Enclosures:

As stated cc w/ enclosures:

See next page THIS DOCUMENT CONTAINS POOR QUAUTY PAGES 80o9000l3%

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Mr. F. L. Clayton, Jr., Senior Vice August 25, 1980 i

President

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Alabama Power Company 1

Post Office Box 2641 l

Birmingham, Alabama 35291 I

Mr. Alan R. Barton.

cc:

Executive Vice President Alabama Power Company Post Office Box 2641 l

Birmingham, Alabama 35291 Mr. Ruble A. Thomas Vice Presideni Southern Company Servicas, Inc.

Post Office Box 2625

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Birmingham, Alabama 35202 f

Mr. George F. Trowbridge Shaw, Pittman, Potts and Trowbridge 1600 M Street, N. W.

Washington, D. C.

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' ENCLOSURE 1

-RE0 VEST FOR ADDITIONAL INFORMATION JOSEPH M. FARLEY NUCLEAR PLANT UNIT 2 DOCKET NO. 50-364 TMI-RELATED REVIEW AREAS Our review of your " Response to the TMI-2 Action Plan" submitted June 20, 1980, nas resulted.in the need for additional information.

Requests and pages-are numbered sequentially. The alpha numeric item designations correspond to the items in the TMI-2 Action Plan. The following requests are included in this enclosure.

Recuest No.

Date Recuested 8

September 8,1980 4

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8.

REACTOR COOLANT SYSTEM VENTS (II.B.1)

Provide your design of the pressurizer vent system per our Novemoer 9,1979 a.

letter! ' Assure that all requirements and recom.endations of that lett'er are satisfied.

b.

If a break in the vent line is beyond that defined as tne smallest LOCA by 10 CFR 50, Appendix A, provide an ECCS performance analysis for a complete spectrum of breaks in the Ent line. Assure that the tnerr.al-hycraulic modeling is consistent with break location.

Demonstrate that the exhaust from the vent system will not impinge en other c.

equipment, that the vent system will vent RCS hot legs, ard that the vent exhaust is to a portion of the containment with maximum ventilation and I

l cooling.

d.

Demonstrate that the vent system is qualified to RPS safety grade standards.

Include seismic design, IEEE-279 requirements, minimize inadvertent actuations, vent valve position indication in the control room, qualification to pass non-concensibles, steam, water and combinations, thereof, etc.

1 Provide procedural guidelines and analytical bases (preferably, generically e.

ceveloped by owr.ers group) for vent operation and termination as related to plant performance. The procedures snould be based on the following :riteria.

(1) the plant mt.st meet'tne requirements of 10 CFR 50.46 and 10 CFR 50.4 for a

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DBA's; and (2) the plants ability to maintain core cooling and containment integrity for events beyond DBAs must be increased. Procedures snould also address methods to (1) assure natural circulation througn the U-tube portions of the steam generator with tne potential accumulation of gases in this region, and (2) assure that combustible limits are not exceeded.

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ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION JOSEPH M. FARLEY NUCLEAR PLANT, UNIT 2 DOCKET NO. 50-364 NON-TMI RELATED REVIEW AREAS Requests from the following branches in NRC are included in this enclosure.

Requests and pages are numbered sequentially with respect to requests transmitted following issuance of SER Supplement No. 3.

BRANCH PAGE NO.

Containment Systems Branch 020-1 Instrumentation and Control Systems 030-1 through 030-4 Branch Materials Engineering Branch 122-3 Reactor Systems Branch 210-6 through 210-13

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020-1 020.0 CONTAINMENT SYSTEMS BRANCH 020.1 Our review of the containment isolation system has also included review of the containment purge system. This system will be used to reduce airborne radioactivity in the containment to permit personnel entry.

In order to complete our review of the purge system, we require the following information:

1) A description of the containment purge system desi.gn_t. hat assures.__

blockage of the purge valves bi 3ebris will not occur. The description should include quality and seismic classification of~ ~ ~

the blockage prevention measures.

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2) A description of the means for detecting high r..ioactivity conditions prior to opening the purge valves.

3) An estimation of the time period that the purge valves will be open during the year with justification for the duration.

In response to item 3 of 020.1, consider the following staff positions.

The staff has recently determined that additional restrictions should be placed upon containment purging during plant operation at plants which are currently authorized to purge as frequently as necessary during plant operation, as Sequoyah is. Additional restrictions on purging will decrease the likelihood of a LOCA occurring while the purge system lines are open. Such open lines constitute a direct connection between the containment atmosphere and the outside enviornment, and failure of the redundant purge system isolation valves to close as required during LOCA, though they may have been properly tested and qualified, would result in offsite doses far in excess of 10 CFR Part 100 guidelines.

Therefore, we require that the Farley plant shall limit use of the containment purge system to a total of no more than 90 hours0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> per year, per reactor unit, during the normal plant operating modes of startup, power, hot standby, and hot shutdown, with only one' pair of purge system lines open at a time. Thus, the 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br /> limit applies to the total time in use of the purge system lines taken all together.

!n the cold shut-down and refueling modes, all purge systems may be used simutaneously and without time limitation. Technical specifications to reflect this requirement should be proposed.

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e 030.0 INSTRUMENTATION & CONTROL-SYSTEMS BRANCH 030.1 POTENTIAL DESIGN DEFICIENCIES IN BYPASS, OVERRIDE, AND RESET CIRCUITS OF ENGINEERED SAFETY FEATURES DISCUSSION OF DEFICIENCIES Several instances have been reported where automatic closure of the containment ventilation / purge valves would not have occurred because the safety actuation signals were either manually overriden or bypassed (blocked) during normal plant operations.

In addition, a related design deficiency with regard to the resetting of engineered safety feature actuation signals has been found at several operating facilities where, upon the reset of an ESF signal, certain safety related equipment would return to its non-safety mode.

Specifically, on June 25, 1978, Northeast Nuclear Energy Company discovered that intermittent containment purge operations had been conducted at Millstone Unit No. 2 with the safety actuation signals to redundant containment purge isolation valves (.48 inch butterfly valves) manually overriden and inoperable. The isola-tion signals which are required to automatically close the purge valves to assure containment integrity were manually overriden to allow purging of containment with a high radiation signal present. The manual override circuitry designed by the plant's architect / engineer defeated not only the high radiation signal but also all other isolation signals to these valves. To manually override a safety actuation signal, the operator cycles the valve control switch to the closed position and then to the open position, This action energized a relay which l

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- l 030-2 blocked the safety signal and allowed manual operation independent of any safety actuation signal. This circuitry wcs designed to permit reopening of certain valves after an accident to allow manual operation of required safety equipment.

On September 8,1978, the staff was advised that, as a matter of routine, Salem Unit No.1 had been venting the containment through the containment ventilation system valves to reduce pressure.

In certain instances this venting has occurred with the containment high particulate radiation monitor isolation signal to the purge and pressure-vacuum relief valves overridden.

The override of this containment isolation signal was accomplished by re-setting the train A and B reset buttons. Under these circumstances, six valves in the containment vent and purge systems could be opened with the radiation isolation signal present. This override was performed after verify-ing that the actual containment particulate levels were acceptable for vent-s ing. The licensee, after further investigation of this practice, detennined that the reset of the particulate radiation monitor alarm also overrides the containment isolation signal to the purge valves such that the purge valves would not have automatically closed on an emergency core cooling sys-tem (ECCS) safety injection signal, A related design deficiency was discovered during a review of system operation following a recent unit trip and subsequent safety injection at North Anna No.1.

Specificit11y, it was found that certain equipment important to safety (for example, control room habitability system dampers) would return to its non-safety mode following the reset of an ESF signal.

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In addition, many utilities do not have safety grade radiation monitors to initiate containment isolation.

SAFETY SIGNIFICANCE The overriding of certain containment ventilation isolation signals could also bypass other safety actuation signals and thus prevent valve closure when the other isolation signals are present. Although such designs may be acceptable, 4

and even necesse,ry, to accomplish certain reactor functions, they are generally unacceptable where they result in the unnecessary bypassing of safety actuation signals. Where such bypassing is also inadvertent, a more serious situation is created especially where there is no bypass indication system to alert the operator.

i Where the resetting of ESF actuation signals, such as safety injection, directly causes equipment important to safety to return to its non-safety mode, protec-t tive actions of the affected systems could be prematurely negated when the associated actuation signal is reset.

Prompt operator action would be required to assure that the necessary equipment is returned to its emergency mode.

i The use of non-safety grade monitor to initiate containment isolation could seriously degrade the reliability of the isolation system.

I STAFF POSITION s

It is our position that, in addition to other applicable criteria, the follow-

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ing should be satisfied for all operating Itcense applications currently under l

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1) The overriding of one type of safety actuation signal (e.g., parti-culate radiation) should not cause the blocking of any other type of safety actuation signal (e.g., iodine radiation, reactor pressure),

2) physical features (e.g., key lock switches) should be provided to en-sure adequate administrative controls.

3) A system level annunciation of the overridden status.should be provided for every safety system impacted when any override is active.

(Seer.G.

1.47).

4) The following' diverse signals should be provided to initiate isolation of the containment purge / ventilation system: containment high radiation, safety injection actuation, and containment high pressure (where con-tainment high pressure is not a portion of safety injection actuation).

5) The instrumentation systems provided to initiate containment purge ventila-tion isolation should be designed and qualified to Class 1E criteria.

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6) The overriding or resetting of the ESF actuation signal.;hould not cause any equipment to change position.

Accordingly, you are requested to review your protection system design to deter-mine its. degree of conformance to these criteria. You should report the results of your review to us by April 15, 1980, describing any departures from the criteria and the corrective actions to be implemented.

Design departures for which no corrective iction.is planned should be justified.

a0verride: The signal is sti11 present, and it is blocked in order to perform a function contrary to the signal.

bReset: The signal has come and gone, and the circuit is being cleared in order to return it to the normal condition.

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122-3 122.0 MATERIALS ENGINEERING BRANCH - COMPONENT INTEGRITY SECTION 122.10 The contaimnent pressure boundary is constructed using materials meeting the requirements of ASME Section III or various B 31.X piping codes; however, our licensing reviews, including that for Farley 2 do not include evaluation of explicit compliance with these codes.

Consequently, to complete our final SER for Farley in response to the July 18, 1980 memorandum D. G. Eisenhut to R. H. Vollmer, we require the following information:

1.

Identification of the fabrication codes (edition and addenda) and the specific paragraphs in these codes that specify the fracture toughness requirements and acceptance criteria (for weldments and base metals).

Codes and code paragraphs should be identified for all materials which constitute part of the containment pressure boundary (e.g., piping penetrations, personnel airlocks, equipment batch).

2.

The materials test data that certify that the fracture toughness acceptance standards have been met for each of the identified materials in the containment pressure boundary.

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210.0, REACTOR SYSTEMS BRANCH 210.2 When shutting down or starting up the plant, certain automatic safety injection signals are blocked to preclude unwanted

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actuation of these systems. Describe the alarms available to alert tha operator to a failure in the primary or secondary

ystem during this phase of operation and the time frame available to mitigate the consequences of such an accident. Justify the time frame available. The following scenarios should be included in your discussion:

a.

In the event of a LOCA following the closure and power lockout of accumulator isolation valves during plant shutdown, what equipment and procedures would mitigate the consequences.

Provide an analysis showing that the consequences of such an event would be less limiting than the design basis LOCA.

b.

Discuss the scenario and consequences of a moderate energy pipe break in the RHR system immediately after initiating RHR operation while shutting down.

Identify instrumentation t

that will provide alarms for operator actions. Describe operator actions required to terminate the event and, using an event time table, show the time available (after receiving the alarm) to perform the action (s*).

Identify the consequences of the occurrences.

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- 21 0.0 REACTOR SYSTEMS BRANCH 210.3 The Regulatory Requirements Review Committee, in a memorandum from E. Case, j

j Comittee Chairman, to L. Gossick, Executive Director for Operations (dated February 16,1978), has approved a new staff position (BTP RSB 5-1) for the Residual Heat Removal System (RHR). The technical requirements i

for your plant are described below.

Please respor:d to these requirements 4

in sufficient detati to enable the staff to review your compliance in an

. expeditious fashi,on. A copy of BTp RSB 501 is attached.

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1.

Provide safety-grade steam generator dump valves, operators, air and power supplies which meet.the single failure criterion (given credit for limited operator action to correct failures).

I 2.

Provide the capability to cooldown to RHR cut-in conditions in less than 36-hours assuming the most limiting single failure and loss of offsite power or show that manual actions ins 'e or outside containment or return to hot standby until the manual actions or maintenance can be i

performed to correct the failure provides an acceptable alternative.

1 3.

Provide the capability to depressurize the reactor coolant system with only safety-grtde systems assuming a single failure and loss of offsite.

power or show that manual actirns inside or outside containment or remaining at hot standbv uti manual actions or repair:: are complete provides an acceptabl' alternative.

4.

Provide the capability for borating with only safety-grade systems assuming a single failure and loss of offsite power or show that manual actions inside or outside containment or remaining at hot standby until manual action or repairs are completed provides an i

acceptable alternative.

5.

Provide the system and component design features necessary for the prototype testing of both the mixing of the added borated water and the cooldown under natural-circulation conditions with and without a single failure of a steam generator atmospheric dump valve. These tests and analyses will be used to obtain information on cooldown times and the corresponding AFW requirements.

6.

Commit to providing specific procedures for cooling down using i

natural circulation and submit a_ sumary of these procedures.

7.

Provide or require a seismic Category I AFW supply for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />

>; Hot Shutdown plus cooldown to the DHR system cut-in based on the longest time (for only onsite or offsite power and assuming the worst single failure), or show that an adequate alternate seismic Category I source will be available.

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o 210-8 BRANCH TECHNICAL P051f10N RSS 5-1 D151CN Rt001RIMENT5 0F THE Rt51fitIAt HE AT RIM 3 VAL SY51t M BACKGROUND CDC 19 states that, "A control room shall te provided from which actions can be taken to operate the nuclear power unit under normal conditions...."

Normal operating conditions include the shutting down of a reactor; therefore, since the residual heal removal (RHR) system is one of several systems involved in the normal shutdown of all reactors, this system must be operable frem the control room.

GDC 34 states that, " Suitable redundance...shall be provided to assure that for onsite electrical power system cperation (assuming offitte power is not av'a11able) and for offsite electrical power system cperation (assu-ing onsite po-er is not available), the system.

t safety function can be secomplished, assuming a single failure."

In most current plant designs the RHR system has a lower design pressure than the reactor coolant system (RC5), is located outside of containment and is part of the emergency core cooling system (ECC5). However, it is possible for the RHR system to have differsnt

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, design characteristics. For example, the RHP. system might have the same design pressure as the RCS, or be located inside of containment. Plants which may have RHR systems that deviate from current designs will be reviewed on a case-by-case basis. The functional, isolation, pressure relief, pus:p protection, aed test rrquires.ents for the RHR systre are included in this position.

4 BRANCH POSITION

,0 A.

Functional Recutrements The system (s) which can be ussJ to take the reactor from normal operating condi-tions to cold shutdcwn" shall satisfy the functienal requirements listed below.

1.

The design shall be such that the reactor can be taken f rom normal operating conditions to cold shutdown" ssirg only safety grade systems. These systems l

shall satisfy General Design Criteria 1 through 5.

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" Processes involved in cooldown are heat removal, depressurtration, flow circulation, and t

reactivity control. The cola shutdown conditten, as described in the Standard Technical Specifications, refers to a seteritical reactor with a reactor coolant temperature no greater than 200*F for a PWR and 212*F for a BhR.

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suitable intirconnections. ica6 oct eet tor., one isol..t uin c e..ta lit ies t o assure tnat f or onsite electric al no.cr system cot ration (assu iinu ot t sits-power is not available) and f or of f site electrical s,o.er systa m operation (assuming ::::ite power is r.:t :.ati: tic) 14 syst:: f or.ction t n be eccomplishec.

assuming a single failure.

3.

The system (s) shall be capable of being operated f rom the cuntrol room =ith either only onsite or only of f site po-er available. In demonstrating that the system can perform its function assuming a single failure, limited operator act' ion outside of the control room would be consicered acceptable if suitably justified.

I 4.

The system (s) shall be capaD1e of bringing the reactor to a cold shutdowt condition, with only of f site or onsite power available, within a reasonable period of time following shutdown, assuming the most limiting single failure.

i B.

RHR System isolation Recuirements

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l The RHR system shall satisfy the isolation requirements listed Delow.

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1.

The following shall be provided in the suction side of the RHR systes to isolate it from the RCS.

l (a) Isolation shall be provided by at least two power-operated valves in series. The valve positions shall be indicated in the control roce.

(b) The valves shall have independent diverse interlocks to prevent the valves from being opened unless the RCS pressure is below the RHR systee i

design pressure. Failure of a power supply shall not cause any valve to change position.

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(c) The valves shall have independent diverse interlocks to protect against i

g one or both valves being open during an RCS increase above the design j

pressure of the RHR system.

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One of the following shall be provided on the discharge side of the RHR system to isolate it from the RCS:

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(a) The valves, position Indicators, and interlocks described in item 1(a) - (c),

l (b) One or more check valves in series with a normally closed power-operated valve. The power-operated valve position shall be indicated in the control room. If the RHR system discharge line is used for an ECCS function, the poweb operated valve is to be cpened upon receipt of a safety injection signal once the reactor coolant pressure has decreased

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below the ECCS design pressure.

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(d) Two check valves in series proviced that there evi-cetiun provisions to ti htnrss and the permit periodic testing of the check valves f or ital g

testing is perfcrmed at lev.t annual!}.

C.

Pressure Relief Reouirements The RHR system shall satisfy the pressure relief requirements listed below.

1.

To protect the RHR system against accidental overpressurization when it is in operation (not isolated from the RCS), pressure relief in the RHR system shall be provided with relieving capacity in accordance with the ASME Boiler and Pressure Vessel Code. The most limiting pressure transient during the i

plant operating condition when the RHR system is not isolated from the RCS shall be considered when selecting the pressure relieving capacity of the RHR system. For example, during shutdown cooling in a Pk1t with no steam bubble in the pressurizer, inadvertent opetration of an additional charging pump or inadvertent cpening of an ECCS accumulator valve should be considered 1.n selection of the design bases.

s 2.

Fluid discharged through the RHW system pressure relief valves must be col-1ected and contained such that 4 stuck open relief valve will not:

a.

Result in flooding of any safety-related equipment.

b.

Reduce the capability of the ECCS below that needed to mitigate the consequences of a postulated LOCA.

c.

Result in a non-isolatable situation in which the water provided to the RCS to maintain the core in a safe condition is discharged outside of the containment.

I 3.

If interlocks are provided to automatically close the isolation valves when j

the RCS pressure exceeds the RHR system design pressure, adequate relief i

capacity shall be provided during the time period while the valves are l

closing.

1 D.

Pump Protection Reouirements 1

The design and operating procedures of arry RHR system shall have provisions to prevent damage to the RHR system pumps due to overheating, cavitation or loss of adequate pump suction fluid.

E.

Test Reovirements The isolation valve operability and Interlock circuits must be designed so as to permit on line testing when operating in the RHR mode. Testability shall meet the requirerents of IEEE Standard 338 and Regulatory Guide 1.22.

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Regulatory Guide 1.63.

Ihr programs fo* 8 * * ~.11 in lue t ests.it w support inc analysis to Ja) confirm that accou te ciairg 0? burateo.ater added priar to or during cooldown can bt achitved onder rat ral tiie.latiuh conditions and permit estimation of the times required to achieve surl. mining, and (b) confirm that the cooldown under. natural circulation conditions can be achieved within the limits specified in the ceergency operating procedures. Comparison with performance of previously tested plants of similar design r.ay be substituted for these tests.

F.

Operational orne dures The operational procedures for bringing the plant from nurmal operating po.er to cold shutdo-n shall be in conformance with negulatory Guide 1.33.

For pressurized water reactors, the operational procedures shall include specific procedures and information required for cooldown under natural circulation conditions.

G.

Auxiliary Feed-ater SerD1y The seismic Category I water supply for the auxiliary feed.ater system for a PWR shall have suf ficient inventory to permit operation at hot shutdown for at least 4 hcurs, follo.ed by cooldown to the conditions permitting operation of the RHR g

system. The inventory needed for cooldown shall be based on the longest cooldown time needed with either only onsite or only offsite power available with an assumed single failure.

H.

Imolementation For the purposes of implementing the requirements for plant heat removal capa-bility for compliance with this position. plants are divioed into the following three classes:

Full compliance with this position for all plants (custom or Class 1 standard) for which CP or PDA applications are docketed on or after January 1, 1978. See Table 1 for possible solutions for full compliance.

Partial implementation of this position for all plants (custom or Class 2 standard) for which CP sr PDA applications are docketed before January 1.1978 and for which an OL issuance is expected on or after January 1, 1979. See Table 1 for recom= ended implemertation for Class 2 plants.

The extent to which the implementation guidance in Table 1 will be Class 3 backfitted for all operating reactors and all other plants (custos or standard) for which issuance of the OL is expected before January 1,1979. will be based on the combined !&E and DOR review of related plant features for eperating reactars, s.

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TA8LE 1 Possible Solutions for full Compliance with STP R58 5-1 and Recommended laplementation for Class 2 Plants Sesign pequirements Processand[5ystem Possible Solu'ign for secoainended impleeentatten fer orCoseenent/

full Compilance -

Class 2 Plants (see foote IL of 81p R58 5-1

1. Functional lacquirement for Longtermcooling[lIHRdropilne/

Provide double drop line (or valves Compilance util not te require 1 if -

in parallell to prevent single valve it can be shown that correction **

laking to Cold Shutdown failure from stopping R68R coe'ing sinale fativre by manual actions-

a. Capability Using Only Safety function. (Note: This requirement InsIde or outside of contaient ae '

Grade Systems in conjunction with meeting ef fects return to hot standby until ranwal

b. C.pability with either only of single failure for long-term actions (or repairs) are foun1 in

.onsite or only offsite power cooling and isolation requirements be acceptable for the indivl+sil and with single failure involve increased number of plant.

(limited action outside CR to independent power suppites and possibly meet 5F) more than four valves.)

C. Reasonable time for cooldown assumino most limiting 5F and only offstte or only onsite power Heat renoval and RC5 circulation rrovide safety-grade dump valves.

Compilance required.

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,during cooldown to cold shutdoun.

operators, and power supply, etc. so (flote Heed SG cooling to maintain that manual action should not be required e

RCS circulation even after RHR5 in after SSE encept to meet single fatture.

N operation when under natural circula-tion)[steamdumpvalves]

Depressurlsation [Fressurizer Provide upgrading and additional Compliance will not t e reo9leM if auntilary spr or power-operated valves to ensure operation of ava-a) dependence on canual actians relief valve litary pressurlier spray using only inside containment after 5tf pe i

safety-grade subsystem aceting sinele

$1ngle f ailure or b) remalnin : et failure. Possible alternative may involve hot standby antil ranual act ioat using pressurlier power-operated relief or repairs are complete are.

valves which have been upgraded. Neet found to be acceptatile im ' ' ' * -

$$[ and single failure without manual ladividual plant.

operation inside containment.

Soration for cold ghetdown [CVCS Provide procedure and uporadleg where same as above.

and boron sampling /

necessary such that boretton to cold shutdown concentration meets the requirements of I. Solution could range from (i) upgrading and adding valves to have both letdown and charging paths safety grade and meet single failure to (2) use of backup i

procedures involving less cost. for example, boration without letdown may be acceptable and eliminate need for upgrading letdown path. Use of ECCS for injection of borated water may also be acceptable. Need surveillance of boren concentration (boronometer and/dr sampilng L mited operator actinn inside or outside d

o containment if Justifind.

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