ML19316A817

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Amend 25 to License DPR-70,revising Requirements Involving Secondary Water Chemistry & ETS Pertaining to Administrative Controls
ML19316A817
Person / Time
Site: Salem 
Issue date: 04/22/1980
From: Schwencer A
Office of Nuclear Reactor Regulation
To:
Shared Package
ML19316A818 List:
References
NUDOCS 8005270579
Download: ML19316A817 (25)
v • d • e


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tk UNITED STATES o,,

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NUCLEAR REGULATORY COMMISSION

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.E WASHINGTON, D. C. 20555

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,0 PUBLIC SERVICE ELECTRIC AND GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 25 License No. DPR-70 1.

The Nuclear Regulatory Comission (the Comission)'has found that:

A.

The applications for amendment by Public Service Electric and Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) dated January 25, 1980 and January 28, 1980 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The istuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

80062'70 Qg

. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License i

No. DPR-70 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 25, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

The license is also amended by the addition of a new paragraph 2.C.(6) that reads as follows:

(6)

The licensee shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation.

This program shall include:

1.

Identification of a sampling schedule for the critical parameters and control points for these parameters; 2.

Identification of the procedures used to measure the values of the critical parameters; 3.

Identification of process sampling points; 4.

Procedure for recording and management of data; 5.

Procedures defining corrective actions for off control point chemistry conditions; and 6.

A procedure identifying (a) the' authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events. required to initiate corrective action.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION I

($$

J.

A. Schwencer, Chief Operating Reactors Branch #1 Division of Operating Reactors

Attachment:

Changes to the Technical Specifications

.Date of Issuance:

April 22,1980

ATTACHMENT TO LICENSE AMENDMENT N0. 25 FACILITY OPERATING LICENSE NO. DPR-70 DOCKET N0. 50-272 Revise Appendix A as follows:

Insert Pages Remove Pages VI VI 3/4 4-25 3/4 4-25 3/4 4-28 3/4 4-28 3/4 7-11 3/4 7-11 3/4 7-12 3/4 7-12 3/4 7-13 3/4 7-13 B3/4 4-7 B3/4 4-7 B3/4 7-3 B3/4 7-3 B3/4 7-4 B3/4 7-4 6-2 6-2 6-3 6-3 6-5 6-5 6-8 6-8 Revise Appendix B as follows:

Remove Page Insert Page 5.2-2 5.2-2

INDEX LIMITING CONDITIONS FOR OPERATION AND aVRVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4. 6.1 PRIMARY CONTAINMENT Co n ta i nme nt I n teg ri ty..................................

3/4 6-1 Co n ta i nm e n t L e a k a g e..................'..................

3/4 6-2 Con ta i nme nt A i r Loc ks..................................

3/4 6-5 Internal Pressure......................................

3/4 6-6 Air Temperature........................................

3/4 6-7 Conta i nment Structural Integri ty.......................

3/4 6-8 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System...............................

3/4 6-9 S p ra y Ad d i t i v e Sys t em..................................

3/4 6-10 Conta i nment Cooli ng 3ystem.............................

3/4 6-11 3/4.6.3 CONTAINMENT ISOLATION VALVES...........................

3/4 6-12 l

3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers.....................................

3/4 6-18 Electric Hydrogen Recombiners..........................-

3/4 6-19 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE Safety Va1ves..........................................

3/4 7-1 Auxiliary Feedwater System.............................

3/4 7-5 Aux i l ia ry Feed Stora g e Ta nk............................

3/4 7-7 Activity...............................................

3/4 7-8 Main Steam Line Isolation Va1ves.......................

3/4 7-10 SALEM - UNIT 1 VI Amendment No. 25

INDEX LIMITINGCONDITIbNSFOROPERATIONANDSURVEILLANCEREQUIREMENTS SECTION Pace 3/4.4.7 CHEMISTRY..............................................

3/4 4-17 3/4.4.8 SPECIFIC ACTIVITY......................................

3/4 4-20 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Rea cto r Cool ant Sys tem.................................

3/4 4-24 Pressurizer............................................

3/4 4-29 Overpressure Protection Systems........................

3/4 4-30 6

3/4.4.10 STRUCTURAL INTEGRITY i

ASME Code Class 1, 2 and 3 Components..................

3/4 4-32 3/4~.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 ACCUMULATORS...........................................

3/4 5-1 avg-350*F.........................

3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - T 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350 F.........................

3/4 5-6

'.3/4.5.4 BORON INJECTION SYSTEM Baron Injection Tank...................................

3/4 5-7 Ileat Tracing...........................................

3/4 5-8 3/4.5.5 REFUELING WATER STORAGE TANK...........................

3/4 5-9 SALEM - UNIT 1 V

Amendment No. 24

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determitied to be within the limits at least once per 30 minutes during system teatup, cooldown, and inservice leek and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiati.on surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H.

The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.

SALEM - UNIT 1.

3/4 4-25

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PLANT SYSTEMS 3/4.7.2 STEAM GENERATOR PRESSU"E/ TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2.1 The temperatures of both the primary and secondary coolants in the steam generators shall be > 70*F when the pressure of either coolant in the steam generator is > 200 psig.

APPLICABILITY: At all times.

i ACTION:

b With the requirements of the above specification not satisfied:

a.

Reduce the steam generator pressure of the applicab'e side to

< 200 psig within 30 minutes, and b.

Perform an engineeririg evaluation to determine the effect of the overpressurization on the structural integrity of the steam generator.

Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200 F.

SURVEILLANCE REQUIREMENTS 4.7.2.1 The pressure in each side of the steam generator shall be determined to be < 200 psig at least once per hour when the temperature of either the primary or secondary coolant is < 70 F.

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O REACTOR COOLANT SYSTEM BASES The reactor vessel materiaJs have been tested to determine their initial RT

the results of these tests are shown in Table B 3/4.4-1.

Reactor opehItion and resultant fast neutron (E>l Mev) irradiation will N

cause an increase in the RT Therefore, an adjusted reference temperature,baseduponthekIu.ence and copper content of the material g

in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2.

The heatup and cooldown limit curves (Figures 3.4-2 and 3.4-3) include at the end of 13 EFPY, as predicted adjustments for this shift in RTg well as adjustments for possible errors in kke pressure and temperature sensing instruments.

periodically during operatiBRT,of the vessel material will be establis The actual shift in RT 3

,,,,y,,,,,3,,,,,,,,,,,,,,cco,g,,c, with ASTM E185-70, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.

Since the neutron spectra at the irradiation samples 3nd vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ART determined from the surveillance capsule is different from the calbhIated ART f r the equivalent capsule radiation exposure.

NDT The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature require-ments of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are in accordance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

SALEM - UNIT 1 B 3/4 4-7 Amendment No. 25

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J h PLANT SYSTEMS BASES 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses. 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within contain-ment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. l SALEM-UNIT 1 B 3/4 7-3 Amendment No. 25 I

o PLANT SYSTEMS BASES 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION ~ The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations t ~ of 70 F and 200 psig are based on average steam generator impact values taken at 10 F and are sufficient to prevent brittle fracture. 3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the canponent cooling water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses. 3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the service water system ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident conditions within acceptable limits. SALEM - UNIT 1 B 3/4 7-4 Amendment No. 25

i 6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Station Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsi- ~ bility during his absence. 6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown on Figure 6.2-1. FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and: Each on duty shift shall be composed of at least the' minimum a. shift crew composition shown in Table 6.2-1. b. At least one licensed Operator shall be in the control room when fuel is in the reactor. c. At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips, d. An individual qualified in radiation protection proc:dures shall be on site when fuel is in the reactor. All CORE ALTERATIONS after the initial fuel loading shall be e. directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation. f. A Fire Brigade of at least 3 members shall be maintained onsite at all times. The Fire Brigade shall not include 4 members of the minimum shift crew necessary for safe shutdown of the unit or any personnel required for other essential functions during a fire emergency. SALEM-UNIT 1 6-1 Amendment No. II Y

r SENIOR g VICE PRESIDENT - 8-ENE RGY SUPPLY & Q ENGINEERING 2 VICE PRESIDENT - VICE PRESIDENT - GINEE tNG PRODUCTION FUE L SUPPLY AND CONSTRUCTION l NUCLEAR REVIEW BOARD GENER AL MANAGER - GENER AL MAN AGE R - GENERAL MANAGER - MAN AGER - GENER AL MAN AGER - LtCENSING & ELECTRIC FUE L SUPf" Y OUALITY ASSURANCE PRODUCTION

  • ENGINEERING ENVIRONMENT l

I M AN AGER - MANAGER - M ANAGE R ' Responsible for Oversti NUCLEAR SALEM GENERATING METHODS N Fire Protection Program. OPERATIONS STATION I ^ TR AINING g O.A. NUCLE AR FUEL HEALTH E NGINE E R ENGINEER CYCLE ENGINEE,R, PHYSICIST g ENG NEER a i rOwin- - - ,l I 1 SENIOR NUCLE AR ASSISTANT I CUR RICULUM TRAINING y i l STATION O.A. I SUPERVISOR ENGINEER l ENGINE ER (SALEMI l i I a - - = ~~. .ts m FIGURE 6.21. OFFSITE ORGANIZATION FOR FACILITY MANAGEMENT AND TECHNICAL SUPPORT STAFF w j thi ( Nx a l

t t i I MANAGER-OUA LIT Y I SALfM l ASSURANCE GE NE R A TING l E NGINE E R l ST ATION

  • r-t 3

reiy L -SITE __] OFF e C SORC Z. t I I ,a MAINTENANCE ASSISTANT CHIE F ENGINE ER TO ENGWE E R T MANAGER I OF FICE ST A TION STATION STATION REACTOR AOMINIS-PERF ORMANCE OPE R ATING, QU ALIT y ENGINE E R p ASS HANCE 5 1 W I I I I y-e SENIOR SE NIOR SENIOR SENIOR SE NIOR SENIOR SHIFT PS RF OR MANCE PE RF ORMANCE PE R F OP'J ANCE MAIN TE N ANCE NUCLEAR SE CURITY SAFETY SUPE RVISOR-SUPf RVISOR-SUPERVISOR-SUPF '.svlSOR. SUPE RVISOR CUR RICULUM SUPE RVISOR SUPE RVISOR SRO l&C R AD PROT. C'.E MISTR Y SUPERVISOR I I I_ l 1 >B 8 ^ MERWSOR WPE WM WPE WISOR W N ISOR ST AF F S E RVISOR-STAFF P OR SRO O-i 3 ED3r+ SRO - SENIOR RE ACTOR OPERATOR BOILER REPAIR CONTROL TECHNICIANS TECHNICIANS MECHANIC gyggy 2 RO - RE ACTOR OPER ATOR OPE R ATORS-llCHNICI ANS NUCLEAR NUCLEAR O T Eithe the Che' E ae.aew ee the Simea Oppeewie Eegwww m'8 RO ELECTRIClANS meet the ees so..we of ANSI NIO 11971 f ee tDe poe.eien of N Opeesteas Measese A's Somer Sheet Seposeneet esse espert te sa S end.d.se who een heid e Seaser Reacter Opeestee's fareaes and heee e.s pene e. a.d.i.m piens ese ace of wheCh e MACHINISTS s awawa ee e.e p. .h n be a.cien pe.= paeae esp =.aee. A y mee wa of e.a et the opmeesag four yens of powee plaat esswe-E 0U1PMENT TECHNICAL S AT y s

i. ace mee ne w.ie,4 b, e...f eci-v ewaemoea e ocase== e' OPE R AT On MSISMN TS C

G e e ,eemed eechaw p ae ea. fu one ewi 6 w h3 Responsible for the Fne Protection Proyem FIGURE 6.2.2 FACILITY ORG ANIZATION - SALEM GENEftATING STATION Nus

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION # LICENSE APPLICABLE MODES CATEGORY 1, 2, 3 & 4 5&6 SOL 1 1* OL 2 1 Non-Licensed 2 1

  • Does not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS after the initial fuel loading.
  1. Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

s SALEM-UNIT 1 6-4

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for the Senior Performance Supervisor - Chemistry /HP who shall meet or exceed the qualifications of Re_gulatory Guide 1.8, September 1975. 6.4 TRAINING 4 6.4.1 A retraining and replacement training program for the facility staff shall be coordinated by the Assistant to Manager and under the i direction of the Training Engineer and shall meet or exceed the require-ments and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Safety Supervisor and shall meet or exceed the requirements of Section 27 of the NFPA Code-1975, except for Fire Brigade training sessions which shall be held at least quarterly. 6.5 REVIEW AND AUDIT 6.5.1 STATION OPERATIONS REVIEW COMMITTEE (SORC) FUNCTION 6.5.1.1 The Station Operations Review Comittee shall function to advise the Station Manager on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The Station Operations Review Committee shall be composed of the: Chairman: Chief Engineer Vice Chairman: Assistant to Manager Member: Station Operating Engineer Member: Station Perfonnance Engineer Member: Reactor Engineer Member: Senior Shift Supervisor Member: Senior Performance Supervisor - I&C Member: Senior Performance Supervisor - Chemistry Member: Senior Performance Supervisor - Rad Protection Member: Senior Maintenance Supervisor Member Maintenance Engineer ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the S0RC Chairman to serve on a temporary basis; however, ro more than two alternates shall participate as voting members in 50RC activities at any one time. SALEM-UNIT 1 6-5 Amendment No. 79. J7, 73,26 L

ADMINISTRATIVE CONTROLS l MEETING FREQUENCY 6.5.1.4 The 50RC shall meet at least once per calendar month and as convened by the SORC Chairman or his designated alternate. QUORUM 6.5.1.5 A quorum of the 50RC shall consist of the Chairman or his designated alternate and four members including alternates. RESPONSIBILITIES 6.5.l.6 The Station Operations Review Cannittee.shall be responsible for: a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) any other proposed procedures or changes thereto as determined by the Station Manager to affect nuclear safety. b. Review of all proposed tests and experiments that affect nuclear safety, c. Review of all proposed changes to Appendix "A" Technical Specifications. d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety. e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the General Manager - Electric Production and to the Chairman of the Nuclear Review Board. f. Review of events requiring 24 hour written notification to the Commission. g. ' Review of facility operations to detect potential nuclear safety hazards. h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Chairman of the Nuclear Review Board. SALEM-UNIT 1 6-6

v ADMINISTRATIVE CONTROLS i. Review of the Plant Security Plan and implenenting' procedures and shall submit recommended changes to the Chairman of the Nuclear Review Board. j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the Nuclear Review Board. AUTHORITY 6.5.1.7 The Station Operations Review Committee shall: a. Recommend to the Station Manager written approval or disapproval of items considered under 6.5.1.6(a) through (d)above. b. Render detenninations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitetes an unreviewed safety question. c. Provide written notification within 24 hours to the General Manager-Electric Production and the Nuclear Review Board of disagreement between the SORC and the Station Manager; however, the Station Manager shall have responsibility for resolution of such disagreements pursuant to 6.1.1 1 above. RECORDS 6.5.1.8 The Station Operations Review Committee shall maintain written minutes of each meeting and copies shall be f.rovided to the General Manager-Electric Production and Chairman of the Nuclear Review Board. 6.5.2 NUCLEAR REVIEW BOARD (NRB) FUNCTION, 6.5.2.1 The Nuclear Review Board shall function to provide independent review and audit of designated activities in the areas of: a. nuclear power plant operations b. nuclear engineering SALEM-UNIT l 6-7 g

ADMINISTRATIVE CONTROLS c. chemistry and radiochemistry d. metallurgy e. instrumentation and-control f. radiological safety 9 mechanical and electrical engineering h. quality assurance practices COMPOSITION 6.5.2.2 The NRB shall be composed of the: Chairman: General Manager-Electric Production Vice Chairman: Assistant to General Manager-Fuel Supply Member: General Manager-Licensing and Environment Member: Manager-Nuclear Operations Member: Manager-Quality Assurance Member: Project Manager-Hope Creek Member: Manager-Salem Generating Station Member: Principal Engineer Member: Manager - Hope Creek Generating Station ALTEitNATES 6.5.2.3 All alternate members shall be appointed in writing by the NRB Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NRB activities at any one time. CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NRB Chairman to provide expert advice to the NRB. MEETING FREQUENCY 6.5.2.5 The NRB shall meet at least once ar calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter. SALEM-UNIT 1 6-8 AmendmentNo.5,9,If,f l

f SENIOR En VICE PRESIDENT - h ENERGY SUPPLY & 2 ENGINEERING b =4 I I ~ VICE PRES 30ENT - VICE PRESIDENT - g, ,, g RODUCUON WEL SWW AND CONSTRUCTION i NUCLEAR l REVIEW BOARD GENE R Al. mar 4 AGER - GENE R AL M AN AGER - GENERAL MANAGER - MANAGER h DOCTION* CEN Surf ENGINEERING OUAllTY ASSURANCE q R MENT l .m MAN AGER - MANAGE R - M ' Responsible for Overell MAN AGE n NUCLEAR SALEM GENERATING Fire Protection Proysm. METHODS OPE R ATIONS STATION I I I I I NUCLEAR TRAINING g O.A. NUCLEAR FUEL HCALTH NN ENGINEER I SICIST ENG NEER CL E,D i I _N SITE O et l I SENIOR NUCLE AR AS$lSTANT 2 l l CURRICULUM TR AINING .O 3 STATION O.A. 1 SUPERVISOR ENGINEER ENGINE ER (SALEMI l 3 I 8 L O! I

=

FIGURE 5.2-lOFFSITE ORGANIZATION FOR FACILITY MANAGEMENT AND TECHNICAL SUPPORT STAFF .}}

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