ML20009D086

Forwards Responses to SER Open Items 36,44,52,59,62 & NUREG-0737 Items II.K.3.28 & Item III.D.1.1

ML20009D086
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	07/21/1981
From:	Mccaffrey B LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.3.28, TASK-3.D.1.1, TASK-TM SNRC-601, NUDOCS 8107230108
Download: ML20009D086 (22)

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1, M/aermearms. SHOREHAM NUCLEAR POWER STATION huumay.mm P.O. DOX 618, NORTH COUNTRY ROAD + WADING RIVER, N.Y.11792 July 21, 1981 SNRC-601

/e Mr. Harold R. Denton Ya E , %

Office of Nuclear Reactor Regulation Q \;'Jg U.S. Nuclear Regulatory Commission 2:n ~gSj! i Washington, D.C. 20555 '

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1> D si SH0REHAM NUCLEAR POWER STATION - Unit 1 Docket No. 50-322 fg l W. \ .S

Dear Mr. Denton:

Enclosed herewith are sixty (60) copies of LILC0 responses to specific NRC concerns which were previously identified as requiring additional information to complete NRC review. Attachment A provides a list of the specific responses included.

If you require additional information or clarification, please do not hesitate to contact this office.

Very truly yours,

.kAW ' y B. R. McCaffr y /

Manager, Project Engineering Shoreham Nuclear Power Station Enclosures cc: J. Higgins QOO$

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/ l 8107230108 810721 PDR ADOCK 05000322 E PDR F C-893 5

% n-Attachment A L:. .-

Additional 1information is provided for the following items:

1) lSER Open' Item No. 36 - Containment Purge System 2)- SER Open Item No. 44 - Level Measurement Errors

.3) SER Open Item No. 52 - Management Organization

4) SER Open Item No. 59 - LILC0 Response to Staff Position Regarding Interim Actions for_ Control of Heavy Loads.

15). SER Open Item Nos. 62 (Reg. Guide 1.58 Rev. 1) and 63 (R.G. 1.146) 6)- NUREG-0737 Item II.K.3.28 - Verify Qualification of Accumulators on Automatic Depressurization System Valves

-. 7)~ NUREG-0737 Item III.D.1.1 ' Primary Co lant Sources Outside the Containment Structure r

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Item No. 36 - Containment Purge System 4

.The drywell'and suppression chamber vent purge valves are six inches in size and 150 pound class. The governing code for design is ASME Section III Code Class 2. The. valves have a design pressure of 48 psig and temperature of 90 F. The ambient temperature specified for the valves' environment is_ The 40 ~120 F.~ The valve body form is globe type with quick opening trim.

valve actuator is~a spring and diaphragm with a 2-inch stroke. Air to ..

the actuator.is 125~psig. Valve actuator operates on 80 psig. In additiofi-to' ASME Section III~ Class 2 design, the. valves have also been seismically qualified (including Mark II. loads) by calculation. Additional seismic

- -evaluation was performed which required analysis to assure the function and

. structural integrity. The valves have been analyzed for stress and deflection

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due to combined seismic loads (simultaneously vertical and horizontal),

-pressure, and maximum operator load, and found that valve function is unimpaired. In addition, the associated solenoid valves are environmentally qualified in' accordance with the requirements of NUREG-0588.

LThese valves were also stroke tested immediately following testing for

' leakage, (hydrostatic test). ~The bench test was performed in a test fixture whir.h did not account for an actual system dcwnstream piping configuration inor could a. constant differential pressure be maintained across the valve during stroke. test.

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~- ' , 5 Shoreham Outstanding SER Issue #44

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Level Measurement Errors-LEVEL MEASUREMENT ERRORS

'(due to environmental- temperature effects on level instrument reference._ ,

legs)-

Reactor vessel' water level is' measured by means of.a produced differentiEi pressure between a-reference leg and a variable leg. The reference leg jis connected to the upper part of the vessel (steam zone) and provides the constant head using an overflow type condensing chamber. The variable

. leg is connected to th+ lower part of the vessel. The produced differential

. pressure'is therefore a function of water level.

General Electric has conducted a reviaw of the effects o_f high drywell

'temperatu ~n reactor vessel water level instrumentation. Although

. instrumei , curacy is affected by varying drywell temperature and boil-off in the reference leg, there would be no impact on the scram or other

' level trip functions, nor would post-accident operator action be impaired.

The worst case scenario evaluated was-as follcws:

o Small_ break LOCA occurs in drywell o Scram and ' auto ECCS (ADS and LPCI/CS) are actuated o Some time after LPCI/ Core Spray have reesta' t'. ished normal RPV water level, operator diverts or shuts off LPCI/ Core

Spray from RPV o 10-12 hours after the initiating event, PRV water level error

.is -at its maximum Upon receipt of low level. alarm (Level 2), operator must

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re-initiate LPCI/ Core Spray injection.

lWere this unlikely series of events to occur it would be necessary for

the operator to restart or .cdirect low pressure ECCS to the PRV in order to avoid core uncos.ery. Shoreham specific analysis verified that

-the operator would have from 10-15 minutes to redirect low pressure ECCS for the worst case scenario presented above, even with a nonconservative

' water level . indication associated with long term boiloff. Operating procedures and training will specifically address the need to be aware of this phenomenon in this small break long term LOCA. Emergency procedure 29.023.09, Reactor Pressure Vessel Flooding, was submitted in SNRC-599 dated July-20,1981, specifically assuring a conservative response to

this phenomenon.

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LThe following statements reflect the resolution 'of the outstanding items associated with SER Open Item No.'52:

.1. Effective July.1,1981 there has been a reorganizationAttached of some of the functions under the office of Vice President Nuclea

-are

. organization.

2.

-In response to the concern over off-site engineering st.pport resourges LILCO will:

.a)

Obtain a Continuing Services Contract with a qualified Architect /

Engineer firm to provide supplementary engineering support, to be in place-prier to Fuel Load, b) .To supplement the technical support of the nuclear organizations a minimum of 10 engineering personnel, assigned to the Corporate Engineering Office, will be designated for nuclear support prior to Fuel Load. . Their first priority will be to respond to the needs of the Shoreham Plant as required.

3. :To further ensure that corporate management decisions adequate a concern for or' consideratici. of matters important to safe o

'of the plant, LILCO will secu. This President Nuclear for matters artecting operation of the plant.

person will have substantial Bh ~over plant operating and manag experience and will be in place' prior to Fuel Load and for a year following Fuel Load or until sud, time as the Vic

4. At and following Fuel Load, a dedicated Comm Security personnel will not be used.

not required to be on shift.

5.'

While LILC0 feels that insufficient credit is being given for the real-time Digital Radiation Monitoring System, we committed Health Physics Technician.

-6. Overtime Policy in compliance with NUREG-0737 (1.A1 and submitted to the NRC.

functions.

7 a)

To clarify our previous position, the Operation Engineer will meet the experience qualif' cations required by ANSI N18.1 - 1971 for An updated resume for J. A. Notaro is the Operations Mana.;er.

attached.

b) Since the Assistant Operations Engineer a ANSI N18.1 - 1971 requirements for the Operations Manager.

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h c) Nuclear Assistant Station Operators (NASO) will meet t e requirements of ANSI.N18.1 - 1971-for License l 8.'

Toisupplement'the operating experience of the operations staff, one ill bli

additional person with ' substantive BWR operating experience4'wThis . role an on site advisor to Senior Plant Management.for. d one

. Manager or. Chief Operating Engineer and their rtrff have accrue

. ant.

sufficient experience in managing an operating

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9. - A general revision to F5AR section 13.2 is under preparation be submitted by December 31, 1981.

L10. with substantive previous BWR operating exp This experienced person will be assigned and shutdown of-a BWR.

to each shift reporting to the Watch Engineer until the plan operating at 100% power. individuals will be submitted as soon as the 7 Item I.B.1.2, LILCO

11. . Supplementing previous responses to NUREG-073 , This group

- commits 1to establishing an On-site Safety Review Group.

will.-be composed of a Chairman and five dedicated multi-disciplin personnel,.the majority of which shall not be recent college grad The Chairman shall report directly to the Manager Nuclear Operations Support and shall transmit formal analyses and Therecommendatio Chairman for presentation to appropriate corporate managshent.

shall- be located off site and shall provide overall direction to the group.

He shall maintain communications with the on-s:+; group through ' egularly scheduled on-site meetings and other informal contacts.

The site.

five dedicated multi-disciplired personnel s coordinate day to day assignments and activities.

of the OSRG shall include:

a)

Assessment of the operating experience of the station and stati of similar design.

b) 1E xamination of appropriate plant operating characteristics a industry /NRC issuances.

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c) .. Review of -plant activities such as maintenance, modification,

, operational problems, and operational analysis. ,

'd) Surveillance of plant operations and maintenance activities tc provide verification that .these. activities are perfor.ned correctly

,'and with minimum human error.

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VICE PRESIDENT NUCLEAR . . .

MANAGER SHOREHAM MANAGER CONSTRUCTION STARTUP AND ENGIh2ERING CONSTR. SC'HEDULE & MANAGER LEAD

" COST SHOREHAM PROJ.

• STARTUP CONTROL ENGINEERING DISCP.

CONTRACTORE FROJECT PROJECT ENGINEERINC LICENSING FIGURE 13.1-4 SHOREHAM PROJECT ORGANIZATION

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SNPS-1 FSAR' J JACK A. NOTAR0 Operating Engineer Long Island Lighting Company Assigned as Operating Engineer of the Shoreham Nuclear Power Station in July,1978.- Responsible for the development and implementation of the Station's operational activities including the direction of day to-~ -

day operation of the unit; startup, operation and shutdown of all station equipment; implementation of initial, requalification, and 2_

-replacement training programs for licensed and unlicensed operators; and development, review, and implementation of the operations section of the Station Operating Manual.

iGraduated fron Brooklyn Technical High School in 1965. Graduated from City College of New York in 1970 with a Bachelor's Degree in Mechanical Eagineering. Received a Masters of Business Administration Degree in

,1974 from Adelphi University.

Completed the General Electric Co. Boiling Water Reactor Simulator Program in July, 1976, and obtained certification as a Senior Reactor

' Operator.

Completed the follewing industry seminars and training programs:

'a) BWR Design Orientation - General Electric Co.

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.b) BWR Technology -

General Electric Co. _.

~ c) BWR Observation Training - General Electric Co.

sd) Nuclear Power Plant Technology - General Physics Corp. ,

c) Radiation Protection - LILCO Evening Institute f) Basic Health Physics - Brookhaven National Laboratory

. g) Vibration Analysis - IRD Mechanalysis, Inc.

h) Statics, Strength of Materials, & Dynamics - LILCO Evening Institute

-1) Management of Maintenance Storekeeping & Inventories - Management

-Dynamics Institute j) QA for the Nuclear Industry' - Fcat-A-Matrix and General Physics Corp.

k). ' Inservice Inspection & QA During Operations - Southwest Research Institute

1) Basic Raffography - Corvair Division of General Dynamics m) _ Magnetic Particle & Liquid Penetrant Testing - Magnaflux Corp.

n)' Basic Ultrasonics - Automation Industries o). Nuclear Powar QA - Long Island Section of AWSC p)- Inservice Inspection Symposium - Mirror Insulation

~ q)- Operations Quality Assurance - Stat-A-Matrix r)~ Fire Fighting Training - Suffolk County Fire Department i

June 1970 - Present

- Employed by the Long Island Lighting Company in the Electric Production Department.

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>Harch'1973 - July'1978

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Assigned to the Shoreham Huclear Power Station in the Quality Assurance

.Section and subsequently promoted to Station Operating Quality Assurance

' Engineer responsible for the Section in July, 1974 Responsibility included initial development of the operational quality ~~~ >

assurance program. Responsible for all aspecto associated with its implementation at the station including reviews, audits, surveillance, eJ.

-inspections, selection and training of personnel, development of procedures and instructions, and the utilization of consultants and contractors. Additional responsibilities included licensing and

. inspection activities associated with the U.S. Nuclear Regulatory Commission and interfacing with external and internal organizations required to implement the operational' quality assurance program.

January '1972 -- March 1973 -

Assigned.to the Electric Production Department Staff. Assigned duties included maintenance scheduling, manpower allocation, equipment testing, station performance analysis and special projects.

June 1970 - January 1972

( Assigned to the Nbintenance Section =in the Northport Power Station.

Assigned duties included assisting 1n outages of both a scheduled and forced nature as well as maintaining plant equipment and systems, and completing special projects.

'A member of the American Society for Quality Control, the Edison Electric Institute - Quality Assurance Task Force (EEI-QATF) and the EEI-QAIF Operations Subcommittee.

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-Item No. 59 - LILC0 Response to Staff Position Regarding Interim Actions for Control of Heavy Loads Thelfollowing requirements will be implemented at Shoreham prior to the

placement i3f new '.uel assemblies in the Reactor Building: +:.

1) Safe losipaths will be defined -in accordance with the guidelines set forth in Section 5.1.1 (1) of NUREG-0612.with the exception that floor markings will be limited to "where practical" due to the inherent radial arid polar pathways traveled by the polar crane.

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2) Procedures.will be developed and implemented per the guidelines set forth in Section 5.1.1 -(2) of NUREG-0612.

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3) Crano operators will be trained, qualified and conduct themselves per .the guidelines set forth in Section 5.1.1 (3) of NUREG-0612.

4) Cranes will be inspected, tested, and maintained in accordance with the guidelines set forth in Section 5.1.1 (6) of NUREG-0612.

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(1) Provide sufficient operator training, handling system design, load handling i instructions, and equipment inspection to assure reliable operation of

-the handling system; and (2) Define safe load travel paths through procedures and operator training so

,. that to. the extent practical heavy loads avoid being carried over or near irradiated fuel or safe shutdown equipment; and

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.(3) Provide mechanical stops or electrical interlocks to prevent movement of

- heavy loads over irradiated fuel or in proximity to equipment associ.ated

with redundant shutdown paths. '

Certain alternative measures may be taken to. compensate for deficiencies in

~(2) and (3)- above, such as the inability to prevent a particular beavy load lfrom being brought over spent fuel (e.g., reactor vessel head). These alterna-tive measures can include: increasing crane reliability by providing dual

' load paths for certain components, increased safety factors, and -increased inspection as discussed in Section .1.6 of this report; restricting crane operations in the spent fuel ~ pool area (PWRs) until fuel has decayed so that off-site releases would be sufficiently low if fuel were damaged; or analyzing the effects of postulated load drops to show that consequences are within acceptable limits. Even if one of these alternative measures is selected, (1) and (2) above should still be satisfied to provide maximum practical defense-in-depth.

The following sections provide guidelines on how the above defense-in-depth approach may be satisfied for various plant areas. Fault trees and associated probabilities were developed and used as described in Bases for Guidelines, Section 5.2 of this report, to evaluate the adequacy of these guidelines and

-to assure a consistent level of protection for the various areas.

5.1.1. General All. plants have overhead handling systems that are used to handle' heavy loads in the area of the' reactor vessel or spent fuel in the spent fuel pool.

Additionally, loads may be handled in other areas where their accidental drop may damage safe shutdown systems. Accordingly, all plants should satisfy each of the following for handling heavy loads that could be brought in proximity

'to or over safe shutdown equipment or irradiated fuel in the spent fuel pool area and in containment (PWRs), in the reactor building (BWRs), and in other

plant areas.

(1) Safe load paths should be defined for the movement of heavy loads to 3': minimize the potential for heavy leads, if dropped, to impact irradiated fuel. in the reactor vessel and in the spent fuel pool, or to impact safe shutdown equipment. The path should follow, to the extent practical, j if structural floor members, beams, etc. , such that if the load is dropped, These load paths

.gf . r the structure is more likely to withstand the impact.

b should. be defined in procedures, shown on equipment layout drawings, and clearly marked on the floor in the area where the load is to be handled.

Deviations from defined load paths should require written alternative procedures approved by the plant safety review committee.

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j (2) _ Procedures 4hould be developed to cover load handling operations for

i. heavy loads that are or could be handled over or in proximity to irradiated a

0 fuel or safe shutdown equipment. At a, minimum,._ procedures should cover j_ A handling of those loads listed in-Tsble. 3-l_of this report; -These TC .gr procedures should include: identifica' tion of required equipment; l' Yy ' inspections and acceptance criteria required before movement of load; the t-j ,  ;

steps and proper sequence to be followed in handling the load; defining the safe _ load path; and other special precautions.

/,(3) 9,

} & Crane operators should,be trainedmqualified and-conduct themselves in N '

x accordance'with 'C'hapter 24 ofANST'B30.2-1976, " Overhead and Gantry " ,'

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(4) -Special lifting devices should satisfy the guidelines of ANSI N14.6-1978, 4 Dtandard for Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (4500 kg) or More for Nuclear Materials." This standard L

. should apply to all special lifting devices which carry heavy loads in areas as defined above. For operating plants certain inspections and t y load tests may be accepted in lieu of certain material requirements in NJ the standard. In addition, the stress. design factor stated in

[9 Section 3.2.1.1 of ANSI NT4.6 should be based on the combined maximum static and dynamic loads that could be imparted on the handling device based on characteristics of the crane which will be used.'* This is in

'th. . lieu of the guideline in Section 3.2.1.1 of ANSI NT4.6 which bases the gd_- stress design-factor on only the weight (static load) of the load and of 2- the intervening components of the special handling device.

(5). Lifting devices that are not specially designed should be installed and

used in accordance with the guidelines of ANSI B30.9-1971, " Slings."

However, in selecting the proper sling, the load used should be the sum E

of the static and maximum dynamic load.* The rating identified on the sling should be in terms of the " static load" which produces the maximum D static and dynamic load. Where this restricts slings to use on only certain cranes, the slings should be clearly marked as to the cranes with which they may be used. '

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• (6)..,.Thecrane-should-be inspected, tested, and maintained in accordance with

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4 Chapter 2-2 of ANSI B30.2-1976, " Overhead and Gantry Cranes," with the exception that tests and inspections should be performed prior to use where it is not practical to me the frequencies of ANSI B30.2 fo.-

periodic inspection and test, or where frequency of crane use is less ,

than the specified inspection a d test frequency (e.g. , the polar crane inside a PWR containment m."

refueling operations, and .. >nh he used every 12 to 18 months during ,

operation.

'" r, lly not accessible during power i ANSI B30.2, howevt.,, alls for certain inspections to be performed daily or mont.11y.  ;

For such cranes having limited usage, tha '

inspections, use;) tests, and maintenance should be performed prior to their &

x For the purpose of selecting the proper sling, loads imposed by the SSE need -

not be' included in the dynamic loads imposed on the sling or lif ting device. '

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SER 0 pen' Item Nos. 62'and 63-

-Safety Evaluation Report, Open Item, No. 62 (R.G.1.58, Rev.1) and No. 63 u-

(R.G.1.146) are associated with Generic letter 81-01, " Qualification of Inspection, Examination, and Testing and Audit Personnel", dated May 4,1981.

As requested in the Generic letter, LILCO commits to the-following:

1. . Regulatory positions C.5, 6, 7, 8, and 10 of Regulatory Guide 1.58, Rev. 1.

-- */ 2. Regulatory Guide 1.146.

These commitments will .be fully implemented by six months after fuel load.

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3 411 1 3.28; VERIFY QUALIFICATION _0F ACCUMULATORS'0N AUTOMATIC DEPRESSURIZATION X  ! SYSTEM VALVES The applicant'should addre_ss' the following issues as _part of the qualification I? - -

program.for the ADS accumulators'and the associated equipment.

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- ?l. :The! applicant should define the number of times the ADS valves must be C.

.A capable,of. cycling using only the individual accumulator inventory and

the>1ength of time these accumulators are required to perform:their ifunction following an' accident.

i w EResponseLto' Item 1:' ,

ThisYinformation was.provided.'in our prev'ious respon d to this item andi in 1 response to FSAR . Question 212.97.

-2. The-criteria for the allowable leakage limits should. be provided'along

with the'. bases for the ciiteria. The leakage' limit should include some

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. 1 margin to account for a-possible increase in leakage resulting from the effects 'of a' harsh environment or a seismic event, unless it can be r demonstrated .that the leak rate.will not increase following an accident.

yThe bases. for the allowable _ leakage criteria- should include a discussion

.' Eof the' basis upon which the margin will' be selected or justification for 6, .

- not applying'a margin to the allowable leakage limit.

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? f 3._1The applicant should commit -to periodic leak thsting of the ADS accumolator

system to assure emergency supply for the~ required number and duration of Lvalve:actuations, as defined in Item 1 above.

x 4 E4. .The applican'. shall- propose technical specifications that specify leak

{ test frequency, allowable leak rate, and the actions to be taken if the leakage limi.t is = exceeded.

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Response to Items 2, 3, and 4 :

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'In crder_ to verify the' leak-tight integrity of the ADS short term accumulator -

system, the applicant will leak test each redundant train at the same frequency

as the plant integrated-leak rate test program. This test should be performed

. ~ such that the static head on the SRV's is not sufficient to unseat them. For

' train A, this test will. be accomplished by closing valves 1PS0*MOV-103A and its adjacent outboard manualiglobe vilve, and IP50*M0V-105A and venting tin

< system piping ~ between 1P50*M0V-103A and its associated outboard globe valve.

-With the minimum normal ~ system pressure at 90 p.sig, the short term accumulator m header, thereby isolated, will- be required to maintain a pressure greater than or equal to 70 psig, as measured on header pressure indicator IP50-PI-116A.

wf .Thisitest will be. performed as a function of time adequate to satisfy the short Eterm ADS requirements. The train B header will be similarly tested.

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• 'If at- the end of. the short term accumulator header leak test, it is found tha

.70ipsig cannot be maintained, each ADS short term accumulator shall have its pressure checked locally to verify it is greater than or equal to 70 psig via a' test gauge. Any short term accumulator system unable to maintain 70 psig.will .be repaired / modified, and retested to verify leak-tight integrity.

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5. ~Since th'e ADS accumulator' system is important to safety, it must meet the.e requirements-.of the GDC 2 and 4. _ The ADS accumulator system, and associat'ed

control c"cuitry, from the ADS valve operator out to and including the accumub.cor system isolation check valve should be seismically and environ-i ' mentally qualified. . Acceptable methods for demonstrating this qualification are. given in SRP Sections 3.9.2, 3.10, L and 3.11, as supplemented by the

. Category I requirements of NUREG-0588.

Response to Item 5:

The selimic and environmental qualification criteria for Shocham are described in FSAR Sections 3.10 and 3.11. In addition, comprehensive description and sta+.us reports for both the seismic and environmental qualification program have been submitted by SNRC-575 dated May 28, 1981, and SNRC-576 dated May 27, 1981 respectively. -As described therein, Class IE electrical equipment is qualified jin accordance with NUREG-0588-Category 2.

The ADS accumulator system is qualified in accordance with the program outlined above and, therefore, meets the requirements of GDC 2 and 4.

6. The applicant will perform a leak test prior to initial operation. The applicant should address the action to be taken if the leakage rate during the pre-operational testing exceeds that established in Item 2 above.

Response to Item 6:

The applicant will perform a leak test prior to initial operation. Should the leakage rate exceed that established in Item 2, the system will be

. repaired / modified as required.

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1 NUREG-0737 Item III.D.1.1 - Primary Coolant Sources Outside the Containment Structure A leakage reduction and control program will be developed to monitor leakage from systems outside containment that could contain radioactive fluids "

during a serious transient or accident. The systems to be included in the immediate and continuing leakage reduction programs are: C -

Core Spray HPCI RHR RCIC Hydrogen Recombiners - Primary Containment Atmospheric Control Reactor Water Clean-Up (RWCU)

Coolant Sampling and Post Accident Sampling Reactor Building Equipment Drains (RBED)

Reactor Building Floor Drains Instrumentation Lines Reactor Building Standby Ventilation System (RBSVS)

Post-accident Monitoring System (PAMS)

The leakage, reduction and control program will consist of regular periodi'c

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visu%1: inspections together with detailed visual inspections and quantitative

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leakage measurem'ents'." Detailed system reviews will be conducted on a once per refuel cycle basis as outlined in Table III.D.1.1-1.

Detailed system walkdowns will be conducted of liquid systems and quantitative leakage measurements will be performed on a once per refueling outage cycle. General visual inspections for leakage or' signs of leakage will be conducted on a more frequent basis by operations personnel on accessible portions of the systems. These insp'e ctions will be conducted with the systems

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pressurized to normal ~or test modes. Any excessive leakage will be recorded and appropriate maintenance work requests generated.

Systems contain'ing gases are to be tested by use of tracer gases (helium, freon, or DOP), by pressure decay testirig or metered makeup tests.

All maintenance work requesa written to correct or investigate leakage on systems included in the 6t agram will be marked as " LEAKAGE RELATED".

This will alert maintenanc. ersonnel to assign high priority to this work and will flag these requests for analysis and record keeping.

The Technical Support Group will administer the leakage reouction and control program. Responsibilities will include the following:

a. Coordinate the once per refueling outage detailed inspections and quantitative leakage measurements, and maintain records generated from the inspection.

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b. Review all leakage related maintenance work requests and maintain a record of status for equipment exhibiting leakage.

This-review will also consider initiating modificatim a to reduce system leakage.

c. -Maintain records and provide a written report annuall, to the '

Plant Manager on the following:

s. 1) Actual system leakage rates as determined by the once J per fuel cycle measurements for each system in the program.

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2) ' Analysis of the data indicating the reason for any high value and identifying corrective action to be taken.

3) Status of pending leakage related maintenance work requests or modifications.

.The Shoreham-leakage reduction and control program will be implemented prior to fuel-load. Before beginning full power operation, LILCO will submit to the NRC staff a report of actual measured leakage rates from all systems include . in the program and all preventative maintenance pe' formed as a direct result of the evaluation of this leakage. The report will also identify leakage criteria to be applied during the' first fuel cycle as the basis .for instituting corrective action in the forrt of preventative maintenance. Prior to the start of the second fuel cycle, LILC0 will revise the criteria as necessary based on the experience gained during Shoreham's first fuel cycle. The revised criteria shall then be used as the' basis for long term leakage mon'~.oring activity at Shoreham.

The following systems which may contain radioactive fluids during a serious transient or accident were not included in the Shoreham leakage reduction and control program for the reasons indicated:

-Suppression Pool Cleanup Mode of fuel Pool Cleanup System -

source: lines from suppression pool automatically isolate upon accident conditions.

-Drywell Equipment and Floor Drains - system isolates automatically upon accident conditions.

-Drywell Air Cooling System - system isolates automatically upon accident conditions.

-RBNVS - system secures upon accident conditions; reactor building ventilation handled by RBSVS which is included in leakage reduction and control program.

In regard to the North Anna and related incidents described in the NRC letter dated 10/17/79. LILC0 is following the recommendations outlined in I&E Circular No. 7s-21, " Prevention of Unplanned Releases of Rddio-

-activity". LILC0 is taking the following preventative measures:

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1. . .' Review of Operating Procedures involving transfer of radioactive pl

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2. - Review of "as-built" systems having the potential of inadvertent

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! 13. . Review oisiveillance Procedures to a'ddress testing of systems -_

i s and temporary piping which could cause an inadvertent release.

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p t- '4. ': Include I&E Circular 79-21 on Required Reading List.

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TABLE III.D.1.,1-1. .

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SUMMARY

OF SNPS-1 LEAKAGE REDUCTION & CONTROL PROGRAM --

SYSTEM TEST METHOD FREQUENCY RESPONSIBLE GROUP

1. . Core Spray ,-Detailed Inspection and Measurement Once per_ refueling cycle Technical Support Visual (Note 1) During Operability Surveillance Operations

2. HPCS  : Detailed Inspectior. and Measurement Once per refueling cycle Visual (note 1) During Operability Surveillance _0perati_ons 3.. RHR Detailed Inspection and Measurement Once.per refueling cycle "

Visual (Note 1) During Operability Surveillance operations

4. Detailed Inspection RCIC Once per refueling cycle - -

Visual (Note 1) During Operability Surveillance Operations

5. Primary Centainment Metered Air Make-up Once per refueling cycle ,

Atmosphere Control - _

Hydrogen Recombiners

6. RWCU Leakage Inspection and Measurement Once per refueling cycle

7. Coolant Sampling & Leakage Inspection and Measurement Once per refueling cycle -

Post Accident Samp1,ing Visual (Note 1) When samples are drawn Radiation / Chemistry

8. RBED Meter for abnormal input rate Daily ' Operations Reactor Building DOP and Freon Testing Once per refueling cyc~le 9- Standby Ventilation .a., (Note 2) _

p. PAMS Leakage Inspection and Measurement Once' per' refueling _ cycle Technical-Support Group t i _

NOTE 1: Accessible portions only 2: Filters only