ML20236E632

Forwards Response to Request for Addl Info Resulting from 890605 Site Visit by R Pedersen Re FSAR Chapter 12.FSAR Changes Initiated & Will Be Incorporated Into Future FSAR Amend

ML20236E632
Person / Time
Site:	Comanche Peak
Issue date:	07/21/1989
From:	William Cahill TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TXX-89476, NUDOCS 8907260033
Download: ML20236E632 (16)

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. Log # TXX-89476 File # 10010 C

C 912.2 1UELECTRIC i

I L WHilsmi J. CaliiR,Jr.

Juit 21. 1969 i

twuria vu rmacnr i

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.U.S. Nuclear. Regulatory Commission j

Attn: Document Control Desk l

Washington, D.C. 20555-1

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

' DOCKET NOS. 50-445 AND 50-446 RESPONSE TO NRC EEQUEST FOR ADDITIONAL INFORMATION ON FSAR CHAPTER 12 1

REF:

Response to information requested as a result of

l June 5, 1989 site visit by the NRR Radiation Protection Reviewer, Mr. Roger Pedersen.

l 1

Gentlemen:

The attached response addresses the questions which arose as a result of the l-site' visit.of Mr. Roger Pedersen on June 5, 1989.

FSAR changes have been

. initiated and will be incorporated into a future FSAR ammendment.

Sincerely, I

i i

19 - -

William J. Cahill, Jr.

I RMR/mgt.

Attachment Enclosures (2)-

i c - Mr. R. D. Martin, Region IV Resident Inspectors, CPSES (3) i Ms.. M. Malloy. NRR-OSP

)

Mr. J. H. Wilson, NRR-OSP l

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i Mr. R. Pedersen, NRR

()O 8907260033 890721 PDR ADOCK 05000445.

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A PNU 1

400 North Olive Street 1,8 B1 DaHas, Texas 75201

' Attachment te TXX-8 W 6 h.S

" July 21,,1989 V

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RESPONSE TO ROGER PEDERSON COMMENTS-- SITE VISIT 5/13/89 1.. ACCESS TO SPENT FUEL TRANSFER TUBE INSIDE CONTAINMENT Noted that current fence and locked gate control is inadequate to prevent unauthorized access and individuals unwittingly being exposed to dangerous

. levels of radiation.

' RESPONSE:

TU Electric considers gates and administrative controls an unacceptable solution to. control of radiation exposure in the area of the spent fuel transfer tube. TU Electric has consulted with other utilities to ascertain shielding requirements.around fuel transfer tubes. Of the 10 plants contacted, all-but one used additional shielding to. reduce the dose rate. The L

one exception installed a caged fence around the entrance to the maze leading to the transfer tube.

Access to the CPSES fuel transfer, tube is obtained by climbing a 12 foot vertical ladder.- Analysis of dose rates in the area indicates that with no additional shielding, dose rates are several hundred rem /hr at the ladder and in excess of 2 rem /hr at the west barrier, 3 ft above the floor, and 1 ft from the containment wall at elevation 811'.

Dose rates are su'stantially higher u

- on. the ladder.and at the fuel transfer tube. Reduction of dose rates to acceptable levels (less than I rem /hr) requires additional shielding.

A design modification is being processed to 1) extend the east and west fenced barriers vertically to prevent unauthorized access.inside the locked barriers, and 2) remove the vertical ladder and add fixed shielding (iron plates) to reduce the dose rates at the gates to well below I rem /hr.

Shielding calculations are based on the expected intensity of gamma rays emitted from a typical spent fuel element residing in the fuel transfer tube at 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after shutdown.

2.

ACCESS TO THE UPPER LEVEL OF THE INCORE INSTRUMENTATION ROOM (RM 11)

Although the room area radiation monitor and local alarm reduces the possibility of an unwitting exposure, the current fence and locked gate does not completely prevent unauthorized access to the room.

. RESPONSE:

A design modification has been initiated to increase the height of the fence to prevent unauthorized access to the room.

.AttachmenttoTXX-8C3 O

,. -l July 21,1989 Page 2 of 5 3.

JNSTRUMENT CALIBRATION FACILITY This facility is not described in the FSAR. The reviewer informed the TV staff-that, consistent with' Reg. Guide 170 and the Standard Review Plan, the radiation protective design features of this facility must be described in the FSAR. This description should cover (but not be ifmited to) those features that. prevent an individual from transversing the radiation beam and being exposed to dose rates in excess of 500 rc1/hr. TU was informed that there is

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some question as to whether this facility meets the special control features required by 10 CFR 20.203-(c)(b)..The reviewer is following up with NMSS on this question. This is a moot point if TV can show that no one can be exposed to the source beam.

RESPONSE

An FSAR change request has been initiated in response to this item. Changes are proposed in Sections 12.2 and 12.5. is a copy of the FSAR change request.

The FSAR change request includes a description of the instrument calibration facility and describes measures developed to ensure no overaposure of personnel to radiation from the radiation beam.

4.

REVISED RADIATION SOURCE DESCRIPTION This item is related to #3 above in that FSAR Amendment 76 substantially revised the table of radioactive sources (Table 12.2-20). TU was' informed that they should insure that the radiation protection features used to control exposure to all radioactive sources are described in the FSAR (consistent with Reg. Guide 1.70 and the SRP).

RESPONSE

The FSAR change request referred to in item 3 references Section 12.5.3.7 of the FSAR which describes the handling and control of all radioactive materials. This FSAR change request also includes a calibration facility-description for the Cs-137 calibration facility.

5.

RADI0 ACTIVE AIRBORNE CONCENTRATIONS The reviewer noted that the TV response to his question concerning airborne radioactivity did not cover the turbine building mezzanine during operation as indicated by their earlier calculation.

[ Question 2.b, below)

NRC Ouestion 2.b:

"The input parameters are listed in Table 12.2-25 sheets 14 and 15 for the turbine building basement and mezzanine, respectively. Comparing these with the results of airborne 1ritium (H3) for these areas (Table 12.2-26 sheets 16 and 17), seems to indicate that increasing the leak rate about l

~~

AttachmenttoTXX-80' O

July 21, 1989 Page 3 of 5 2.5 times (in rooms of about the same size and ventilation flow rate) increases the H3 airborne concentration by 37 orders of magnitude. Also, L

, justify a design that has.5 MPC of H3 on the turbine building mezzanine during normal' operation."

RESPONSE

TV Electric has previously submitted average airborne radioactivity concentrations in the Turbine Building (TXX-89221, April 28, 1989). A brief analysis is provided below to compare the Turbine Building Basement and l

Mezzanine tritium concentrations. The basic assumptions and methodology in this comparison are consistent with the previous submittal and NUREG-0017, Revision 1.

Discussion and Assumptions:

The Turbine Building ventilation System is designed to direct the incoming supply air to the basement elevations, and to draw building exhaust from the mezzanine elevations. This arrangement produces a general upward air flow such that the " exhaust" from the basement is drawn through the mezzanine elevations prior to being released to the atmosphere. Airborne activity from the basement is, therefore, drawn through various grating and structural openings into the mezzanine elevations, and should be considered to be present in both regions.

For purposes of this analysis, tritium is conservatively treated as a noble gas.

For determination of noble gas releases, NUREG-0017, Revision 1 assumes a release rate into the affected area equal to 1700 lbs/hr of steam activity.

A review of the Design Basis Documents pertaining to the Main Steam, Reheat, Steam Dump, Extraction Steam, Auxiliary Steam, Feedwater and Heater Drains Systems indicates that essentially all of the leakage sources for steam are located on the mezzanine elevations. Additional leakage sources for airborne radioactivity in the basement elevations include evaporation from the Turbine Building Sumps, and a small number of potential steam leakage paths. However, the steam contained in equipment located on the basement elevations (condensate and heater drain systems) is generally at much lower temperatures and pressures than that located on the mezzanine elevations in the more significant steam carrying equipment.

It is conservatively assumed that 25% of the total steam leakage is released to the basement atmosphere. Also, since the airborne activity from the basement is drawn through the mezzanine elevations,100% of the steam leakage is assumed to be released to the mezzanine elevations. These assumptions conservatively include contributions from all sources discussed above.

Basic Data:

Gross Turbine Building Volumes:

Basement 1.72E+6 ft3 Mezzanine 1.77E+6 ft3

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TXX-89476 COM6ergy coff Page 4 of 5 Ventilation Rates:

Basement 2.76E+5 cfm 9.63 air changes per hour

=

Mezzanine 2.76E+5 cfm 9.36 air changes per hour

=

3 Isotopic Data for H -

Half life (T) 12.3 years Decay Constant 6.43E-6/ hour Steam Concentration 0.001 uCi/ gram Maximum Permissible Concentration 5.0E-6. uCi/cc Calculation:

The airborne tritium concentrations are calculated below by using a modified form of the equilibrium concentration equation discussed in FSAR Section 12.2.2.4, assuming the decontamination and partition factors are unity.

C - (K) (LR) (C1) / (Z)

Where:

C - The equilibrium airborne concentration in the affected volume, pCi/cc.

K - 0.016, for units conversion, gram-ft )/lb-cc 3

LR - Effective leak rate into the affected volume, lbs/hr.

Ci - Activity concentration of tritium in steam generator steam, uti/ gram.

Z = V (x )

3 V - Volume of the particular region, ft 3 - The sum of the radioactive decay constant and the ventilation rate,1/hr.

Substituting the appropriate data into the equilibrium equation above yields:

C - (0.016)(0.25*1700)(0.001)

(1.72E+6)(9.63 + 6.43E-6)

C - 4.llE-10 gri/cc, or 8.2E-5 MPC for the Turbine Building Basement, and

Attachment to TXX-8 N D

V

'd-July 21,1989

-Page 5 of:5-C - (0.016)(1.0*1700)(0.001)

'(1.77E+6)(9.36 + 6.43E-6)

C - l'.64E-9 p Ci/cc, or 3.3E-4 MPC for the Turbine Building Mezzanine elevations.

Summary:

The radioactive airborne concentrations in the basement and mezzanine levels of the Turbine Building are comparable as expected, based upon the ventilation rates, release rates, and volumes of the two regions. Also, these values, which are in good agreement with previously calculated values for average Turbine Building concentrations, are small fractions of the MPC.

The previous calculations of plant airborne activities are currently being revised and will include analysis of the airborne' concentrations in the Turbine Building basement and mezzas;ine area.

FSAR Tables 12.2-25 and 12.2-26 will be revised accordingly and incorporated into a future FSAR amendment.

Calculations will be approved and available for review on site after August 31, 1989.

6.

START-UP TESTING RADIATION SHIELDING SURVEYS TU was informed that the NRC staff considered the procedures and programs outlined in the industry consensus standard ANSI /ANS 6.3.1 - 1980 as an adequate survey program.

RESPONSE

A procedure specifying the requirements for the performance of the shield wall surveys at various power levels during-initial startup is in development.

ANSI 6.3.1 will be used as a guideline.

7.

H. P. ORGANIZATION The reviewer provided proposed wording for a revision to their draft Tech Spec that would address the Assistant RPM and their current organization structure.

RESPONSE: is a copy of the proposed change to CPSES Technical Specifications to insure that the individual responsible for the technical aspects of the radiation protection program meets the requirements of Reg. Guide 1.8.

. Enclosure 1 Vo / Jct 9974 "I

,s FIGURE 7.2 UCENSING 00CUMENT CKANCE REGUEST I

NE. 27261 Dd.@ g #

Date:

To:

July }A,8j9g6 4

LDCR NO.:

g LICENSING DOCUMENT CHANGE REQUEST References section er FSAR:

Sections 12.2.1.4 and 12.5.2.1 Descripteen er Change (Markup ettecheek Addition to identify special control of 5,000 ci Cs-137 calibration source. Revision to include detailed description of the Calibration Facility and shielding of the 5,000 ci source.

References:

1) Site visit'of NRR R.P. Branch Reviewer, Roger Pederson, on May 31, 1959

2) TU Electric Memorandum TUS-89719 (Jime 5. 1989)

Jusiericeison:

The additional information was requested by the NRR Radiation Protection Branch Reviewer.

MW'O JOI 17 NO Orgentastion Originating Reevest:

Comanche Peak Engineering DeSCIPUNE NAME INtflALS j GATE Cagnesent Engineer J. K. Warkentin (EIS) M 7[/f/99 Area Superveser L. A. Voicik (E15)

Nh 7//3/ff

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Unit Meneger F. W. Madden (E15)

N,VM' 7.</ r

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v Mechenscot Section Manager F. W. Madden (E15)

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.R,'- r" Electricot Secten Manager yfg Cave Section Manager Lses utenems Engineer

. RECOMMENs Approvat O Otese,revei A reve. O sine A reve.

Sy Deiet Sys Detet uttusess teessette DOCet? Sm'Eev50s Change Schedules For Amendment /Reviseen:

cc: ARMS Egg J.T. Conly (1L,1A) (SW1)

E.J. Schmitt(lL,lA)(005)I R.P. Baker (1L,1A) (CL1)

J.J. Kelley (1L,1A) (009)

D.L. Davis (1L,1A)(006) g J.W. Beck (1L,1A) (ST) 0.W. Lowe (1L,1A) (E07)

D.C. Kay (1L,1A)(ST)

H.D. Bruner (1L,1A) (E21)

A.B. Scott (1L,1A) (010)

ECE 4.53 eEv. O AG 7.

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ATTACHMENT 8.5 (Example) i COMANCHE PEAK STEAM ELECTRIC STATION Tracking Number:

Lieensing Documene Change Requesr t.OrJ2. Q yD %Og'y Document Identification:

FSAR Referenced Section:

12.2.1.4 and 12.5.2.1 Description of Change:

(1) Revised Section 12.2.1.4 to add a brief discussion of controls of radioactive sources.

(2)- Revised Section 12.5.2.1 to add a more detailed description of the Radiation Protection Calibration Facility and the 5000 Ci calibra-tion well source, including the well assembly, shield design, physical barriers, and administrative and procedural controls.

References:

(1) Site visit of NRR Radiation Protection Branch Reviewer, Mr. Roger Pederson, on May 31, 1989.

(2) TU Electric Office Memorandum (LOG #TUS-89719).

Justification:

(

The additional information was requested by the NRR Radiation Protection Branch Reviewer.

l Organization Originating Request: Radiation Protection ov 1 Recommended B Originator:

C Kay/ E Date:

ggg Date:,]jg,g,9,p Lead Lig sing Engineer:

.m Recomme.% Approvai _ Disapproval _

Approved Not Approved _

By:

Date:

By:

Date:

Licensing Engineer Docket Licensing Manager Change Scheduled for Amendment / Revision NE0 4.03 Rev.: 3 Date:

03/07/89 Page 8.8 - 1 of 1

CPSES/FSAR.

O 12.2.1.3.6 GAP ACTIVITY RELEASE l66

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The.gapactivityisthefractionofcoreactivitythatdiffusestothel66 gap between the fuel and cladding. For design basis accidents, the l

noble gas and iodine inventory in the fuel gas is assumed to be 10 l

percent of th'e core noble gas and iodine inventory, except for Kr-85 l

which is assumed to be 30 percent. The design basis gap activities

-l are presented in Table 15.6-8.

l 12.2.1.4 Bvoroduct. Source and Snecial Nuclear Materials Information on the projected types, quantities, forms and uses of byproduct, source and special nuclear material as defined in 10 CFR Part 30,10 CFR Part 40 and 10 CFR Part 70 is provided in Table 12.2-20.

> NSER.T A 12.2.1.5

. Ass - tions. Calculation Me+ W s and Desian Basis l66 for Contained Radiation Sources l 16

(

Radiation shielding'is designed to provide radiation protection for plant personnel during plant operation at maximum calculated thermal power and to limit the normal operation radiation levels at the exclusion area boundary to below those levels allowed for continuous nonoccupational exposure. The plant is capable of continued safe operation with fuel cladding defects in the equivalent to one percent of the fuel rods.

In addition, the shielding provided ensures that, in the event of a hypothetical accident, the integrated offsite' exposure due to the contained activity does not result in any harmful offsite radiation exposures.

k Amendment 66 12.2-11 January 15, 1988

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INSERT A' g

(Addition to Section 12.2.1.4)

Speeee.l AittKfaen is 9Ivsn 'tb ir ;: rr:1, These rad active materials are handled in accordance with the controls described in S tion 12.5.3.7, " Radioactive Materials Control".

rticul

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the 5000 C1 Cs-137 source used for calibration of

^'

pertable radiation detection instrumentation, including high range instrumentation available for performing surveys during emergency conditions. This source is located in the Radiation Protection Calibration Facility and is housed in a Calibration Well Assembly. The

- calibration facility, well assembly, shield design, physical barriers and administrative and procedural controls associated with this source are discussed in Section 12.5.2.1.

i

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CPSES/FSAR 76

'12.5.2 EQUIPMENT, INSTRUMENTATION, AND FACILITIES 76 Radiation Protection will procure and maintain the appropriate monitoring instrumentation and protective equipment and supplies necessary fo_r operating and maintaining CPSES. Additionally, adequate facilities will be available for the conduct of radiation protection activities such as controlling access to Radiation Controlled Areas, locating and storing equipment, perforn.ing calibrations, etc. The following sections provide informat on on the i

types of equipment availaW, including driteria for selectio.1, and radiation protection facilities.

76 12.5.2.1 Radiation Detection Instrumentation and Calibration 76 Facilities 76 Portable, non-portable and laboratory radiation detection equipment are selected to provide the appropriate detection capabilities, ranges, sensitivities, and accuracies required for the anticipated types and levels of radiation to be found at CPSES, whether during normal operations and maintenance, or emergency conditions. Non-portable radiation detection / measuring systems are listed in Table 12.5-1.

These systems are under the control of the Chemistry and Environmental, or Radiation Protection Sections. Personnel will perform functional checks on each counting system in accordance with procedures or instructions fn order to detenmine such things as instrument response, background count rates and counting efficiencies. Records will be maintained for each instrument or counting system. Repair and maintenance of this equipment will be performed by station or contract personnel.

Portable radiation detection equipment is listed in Table 12.5-2.

The equipment listed in Tables 12.5-1 and 12.5-2 provides examples of minimum instrumentation which may be substituted for or replaced by functionally equivalent instrumentation to support operation of the plant. Storage locations for radiation protection instrumentation, such as portable survey instruments, will depend on the function and frequency of use of the instrun,ent.

In general, such instrumentation

-C::

Amendment 76 12.5-4 May 1, 1989

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CPSES/FSAR will be stored at locations that offer convenient accessibility to the 76 work areas, such as the access control office.

Calibration of all' portable and non-portable radiation protection 76-equipment wi.11 be performed by Radiation Protection or Chemistry and Environmental personnel except when other station or contract personnel are required to calibrate certain items such as magnehelic gauges, flow meters, pulse generators, electrometers, or other items

.that are beyond the scope of Radiation Protection or Chemistry and Environmental personnel. Calibration instructions will be written with detailed records of calibration and maintenance of each 56 instrument maintained at the station.

Calibration of portable radiation protection equipment will be 76 performed in the Radiation Protection Calibration Facility (or other suitable location) or by a qualified vendor. The Radiation Protection Calibration Facility is located in the Turbine Building Annex adjacent to the Unit 1 Turbine Building.

This facility

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provides a low background environment, a storage area for calibration 56 sources, adequate work space, and a convenient location.

Calibrations will be performed using radioactive sources traceable to 76 the National Bureau of Standards (N.B.S.) or usi1g transfer instruments, such as electrometers, which have been calibrated using NBS traceable sources.

Each instrument will be labeled with a calibration sticker after it has been successfully calibrated.

> INSERT S 12.5.2.2 Radiation Protection and Access Control Facilities 76 The Radiation Protection and Access control facilities are located in 76 the Unit I side of the Turbine Building on Elevation 810'-6".

Figure 12.5-1 should be referenced for details concerning this area. This 76 area provides office space for Radiation Protection personnel and is arranged to maintain positive control for entry and exit of personnel and equipment to and from the Radiation Controlled areas (RCA).

(

12.5-5 Amendment 76 j

May 1, 1989

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INSERT B (Addition to Section 12.5.2.1)

The Radiation Protection Calibration Facility includes a high activity Cs-137 source (initially about 5,000 Ci in 1983) housed in a calibration L

well assembly. The well assembly provides shielding for the source and I

consists of a 30 foot (below grade) vertical well containing the calibration source which rests on an elevator assembly. The well assembly is designed to ensure that radiation exposures to individuals in unrestricted areas are in compliance with 10CFR Part 20. An elevator operating assembly, consisting of a chain and sprocket assembly with braking ability and a device to indicate the position of the source, is used'to raise, lower or hold the vertical. position of the well source elevator assembly. The calibration well assembly also includes five moveable shields mounted in traeb that are used for establishing desired exposure rates. These shields are locked to shield the active' area above the calibration well during periods when the Calibration Well Source is

, " not"i:'a rili d.

Each shield is also equipped with a handle which allows the operator to move the shields along their tracks without extending any portion of the body into or over the active area of the Calibration Well.

During periods when the source is in use, a high radiation area exists on the roof of the calibration facility. This area is fenced to prevent access, posted with appropriate warning signs and equipped with a warning light that is energized when the source is in use. There is no normal or routine access to the roof of the facility.

The following physical and administrative controls are used to prevent inadvertent exposures resulting from the Calibration Well Source:

1.

During periods when the Calibration Facility'is not attended, the source is lowered to its bottom position, and the physical barriers designed to prevent access to or operation of the Calibration Well Source are locked. These physical barriers include locking the elevator assembly hand wheel, locking the moveable shields, and locking the entrance to the well source room.

2.

Warning lights are mounted on the roof of the calibration facility and in the well source room. These lights are activated by any movement of the moveable shields from their stored position or whenever the well source is raised off the bottom of the calibration well.

3.

The well source assembly contains an upper limit stop which prevents the source from being raised above a designated height that would result in excessive exposures to the operator and in areas outside of the calibration facility.

LO rJ z.

E 4.

All operations'of t3e Calibration Well Source are performed in accordance with approved station procedures.

5.

All activities involving use of.the Calibration Well source require the issuance' of a Radiation Work Permit.

6. ~ Use of the Calibration Well Source.is only allowed by personnel specifically trained and qualified in its-use.

.--_.-.-__.__-.--.________.___-a

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C(N ACHE PEAK STEAM ELECTRIC STATION Tracking Number:

Licensing D:cument Change Request o-Document Identification: cestES Wi T.,2,u.l. spqcs.cu a,

' Referenced Section: f,. 3 Description of Change:

kolclSp&~f5&

h lbe 0 2lab men 8 Ryea

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See Me es%cAed Mdd y P)*

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References:

CPsES Q-t T/5 h.Ef TuS- &l 119, da+ect & / c5 / 89; EG.l.8/

F3AR. I3.1.'A Justification:

hrossides 'asswrence, bh be I E Pree.h b W

gl.tdamce. A R. G. -l. E a,. 4e. e.p g ;.,, p ;),g., & c A.

-ftchwihlespcks *SSe itadia. Mon piechen progre.

-Organization Originating Request:

g g / L. I Originator: S l b e II Approval Recommended by:

Date: (,/gs/si Date:

Lead Licensing Engir.wr:

Recommend-Approval

-l[l Disapproval l[l Approved l[lNotApprovedl[l By:

Date By:

Date:

Licensing Engineer Docket Licensing Manager Change Scheduled for Amendment / Revision 8/23/88 4

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'kDMINISTRATIVECONTROLS f)li]l I

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i UNIT STAFF QUALIFICATIONS u l}

l 6.3 Each member o'i the unit staff shall meet or exceed the min F

l t for for comparable positions, excep j

i 6.3.1 qualifications of ANSI-N18.1-1971 d the 4

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the Radiation Protection Manager", who shall meet or excee I'

l],

for a qualifications of Regulatory Guide 1.8, September 1975,The licensed Operators j.

F I Operators shall also meet or exceed the minimum qualifications of the Radiation Protection Manager.

1 supplemental requirements specified in Sections A and C of Enclosure id 28, 1980 NRC letter to all licensees.

(Prior to meeting 1

technicians, and maintenance k(s) for q!

of the March the qualifications of ANSI-N18.1-1971, personnel may be perm

)

which qualification has been demonstrated.

6_. 4 TRAINING C

A retraining and replacement training program for the unit staff shall be maintained under the direction of the Vice President, il 6.4.1 d

Nuclear Operations and shall meet or exceed the requirements an i

recommendations of ANSI-N18.1-1971, 10CFR55, and shall include familiarization with relevant industry operational experience.

i J f, 6.5 REVIEW AND AUDIT 4

i STATION OPERATIONS REVIEW COMMITTEE (50RC) i 6.5.1 l

FUNCTION The 50RC shall function to advise the Vice President, Nuclear Operations, on all matters related to nuclear safety.

6.5.1.1 I

C0f4 POSITION D

The 50RC shall be composed of managers or individuals d meet reporti~ng directly to managers from the areas listed below anSection 4.2 or 4.

]

I 6.5.1.2 i d the requirements of ANSI N18.1-1971 n

experience.

l 1

I Operations

,l Maintenance j

Instrumentation and Control i

Technical Support l

Radiation Protection Ouality Assurance ggg',4 j

1 Emergency Planning Security 934*now h%9 lC Testing Until the Radiation Protection Manager me ts all qualifications

..';. M who meets per Regulatory Guide 1.8, September 1975,4all of tho 4

":d !:t i r

, atc:ti: ":::; g 4 Mc Radiah REVISI~0NE i

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COMANCHE PEAK - UNIT 1 f

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