ML20245G878

Reactor Facility Annual Rept 1988

ML20245G878
Person / Time
Site:	Armed Forces Radiobiology Research Institute
Issue date:	12/31/1988
From:	Felty J, Talkington G, Ting W DEFENSE, DEPT. OF, DEFENSE NUCLEAR AGENCY
To:
Shared Package
ML20245G854	List: ML20245G850 ML20245G878
References
NUDOCS 8905030255
Download: ML20245G878 (221)

Text

l f , - -

Reactor Facility Annual Report

~~

1 Jan 1988 -

to 31 Dec 1988

L l :::::: l I lh i :1:,

M. L. Moore .

Reactor Facility Director .

....e aa a a a a 0 -

DEFENSE NUCLEAR AGENCY ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE BETHESDA, M ARYLAND 20814-5145 Distribution lin sted to U. S vernment agencies on for official /operat onal use. Other uests for this document must be R

f ii hf70 PDC

ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE TRIGA MARK-F REACTOR FACILITY i I ANNUAL REPORT I

1 JANUARY 1988 to 31 DECEMBER 1988 Pt 3 pared by:

Wendy Ting MAJ James R. Felty

'l SFC Gary F. Talkington

< g I

l Mark Moore OM Reactor Facility Director i I

I A d for Releas : ,

George W.

1. Mv ing, III N'A -

//

Colonel, S F,BSC )

FRRI I

Director, {

I l

1  :

1 I

l t_

,1 1988 A.NNU AL REl> ORT

[J TABLE OF CONTENTS e

Introduction - 3 i

General Information 5 i

.' Sectio'n 1 7 Changes , to the facility design, performance  ;

i- characteristics, operating procedures, and results of surveillance testing i

Section II 13 Energy generated- by current reactor core l i

Section -III . 13  !

Unscheduled shutdowns

'Section IV 14 Safety related corrective maintenance Section V 17 Facility changes, . changes to '. procedures, and l new experiments Section - VI - 17 Summary of safety evaluation changes not submitted to- NRC pursuant to 10 CFR 50.59 Section VII 18 Summary of radioactive effluent released Section VIII 18 Environmental radiological surveys Section IX- 19 Exposures greater than 25% of 10 CFR 20 limits Attachment A Charter for the Armed -Forces Radiobiology Research Institute Reactor and Radiation Facility Safety Committee Attachment B Current Reactor Operating Procedures L

l AFRRI TRIGA Reactor Facility Page 1 l

(1

1988 ANNUAL REPORT t

Attachment C Safety Analysis of Modifications to Upgrade the Reactor Facility at the Armed Forces Radiobiology Research Institute Attachment D License Event Reports Attachment E RRFSC Approval of Special Reactor Authorization l-i l

I .

t 1

i 1

4 1

l-l l

l I

i, I i

I 4

AFRRI TRIGA Reactor Facility Page 2 l l

l 1

I 1988 AS.W AL REPORT I Introduction 1988 began with expectations that the new facility instrumentation would be placed in an operational status. The new microprocessor based instrumentation and control system, developed by General Atomics, arrived at AFRRI and began undergoing pre-installation check out and testing. The year included some notes of frustration, however. Continuing software problems prevented the installation of the new console.

I In October 1988 a concerned employee raised allegations of violations of procedures, technical specifications, parts of the Code of Federal Regulations, and other items not related to reactor operations. An internal invi stigation I by the Defense Nuclear Agency inspector General (DN AIG) found that the majority of the allegations were unsubstantiated and were based on differences of interpretation and understanding of rules and regulations I between the cancerned employee and management. However, based on a recommendation made by the DNAIG team, a thorough, exhaustive review of all reactor operational procedures was conducted, resulting in many changes that greatly benefitted the facility. Because of safety concerns, the employee was temporarily removed from the reactor area. Following a complaint to the Department of Labor (dol) by this employee, the Dot found that management had discriminated against this employee. In a response to the dol, I management disagreed with the dol determination, but complied with the Dot recommendation and the employee was returned to the reactor staff in December 1988. Management still had serious safety concerns and engaged an I industrial psychologist to work with the reactor staff members to ease tensions and improve communication.

)

I The Reactor facility was inspected by the U. S. Nuclear Regulatory Commission (USNRC) during 26-28 Oct 1988 and 7 Nov 1988 (USNRC Report 50-170/88-04).

The results of this inspection coupled with the findings by the Reactor Facility Director (RFD) during a subsequent internal audit resulted in the RFD I placing the reactor in a non-operational status on 16 Dec 1988. This non-operational status allowed time for an intense review and modification of existing procedures as well as time for a thorough procedural training exercise for all SROs.

AFRRI hosted a very successful National meeting of the TRIGA Owners and Users Association in early spring.

Changes made to the reactor facility necessitated the development of documents which described each modification and its applicability under 10 I CFR part 50.59. These included the new microprocessor based instrumentation and control system and several other facility modifications. Each of these changes was supported by the required safety review process. These changes are elaborated upon in Section V of this re' port.

AFRRI TRIGA Reactor Facility Page 3 I

1988 ANNt!AL REPOffl One Senior Reactor Operator (SRO) was added to the staff during CY 1988 along with two trainees. One SRO departed during the year. Requests from non-AFRRI investigators to use the reactor facility continued to supplement the substantial in-house experimental workload. These experimenters included representatives from the National Institutes of Health (NIH), Federal Bureau of l Investigation (FBI), and the National Institute of Standards and Technology l

(NIST). Again this year the reactor staff was tasked to provide personnel to assist in conducting Department of the Army Inspector General (DAIG) inspections of the Fast Burst Reactor at White Sands Missile Range, New Mexico and the Fast Burst Reactor at Aberdeen Proving Grounds, Maryland.

A new charter for the Reactor and Radiation Facility Safety Committee (RRFSC) was approved. This charter, as shown in Attachment A to this report, details the functions and responsibilities of the RRFSC.

Two License Event Reports (LER) were submitted to the USNRC during the calender year. Positive steps were implemented to prevent further occurrences of the events detailed in these LERs in Attachment D to this re por t.

The remainder of this report is written in a format to include notification items required by the AFRRI TRIGA Reactor Technical Specifications. Items not specifically required but of a general informational value are presented in the General Information section. Each section following the general information corresponds to the required section as listed in Section 6.6.1.b of the AFRRI TRIGA Reactor Technical Specifications.

1 i

l i

)

AFRRI TRIGA Reactor Facility Page 4 l l

l l

1988 ANNU AL REPOHI General Information l All personnel listed held their positions as listed throughout the entire year unless otherwise specified.

1. Current key AFRRI personnel (as of 31 Dec 1988) are as follows:

f Director - Col George W. Irving, III, BSC, USAF l

Scientific Director - Capt Richard Walker, MSC, USN (position vacant since 8 Aug 1988)

Chairman, Radiation Sources Department - Mr. Mark Moore (SRO)

Manager, Radiation Sources Program - MAJ Leonard A. Alt (SRO) i Chairman, Safety and Health Department - Mr. Douglas Ashby (effective 4 Dec 1988)

2. Current key Reactor Operations Personnel:

Reactor Facility Director - Mr. Mark Moore (SRO)

Chief, Reactor Division - MAJ James R. Felty (SRO)

Reactor Operations Supervisor - Capt Kenneth Hodgdon (SRO)

(appointment terminated 24 Aug 1988)

Reactor Operations Supervisor - MAJ James R. Felty (SRO)

(appointment effective 24 Aug 1988)

Asst Reactor Division Chief - Ms Wendy Ting (SRO, effective 23 Aug 1988)

Training Coordinator - SFC Gary F. Talkington (SRO)

Procurement Coordinator - SFC Philip Cartwright (SRO)

Maintenance Coordinator - SFC Wayne Reed (SRO)

Nuclear Engineer - Ms Angela Munno (SRO)

Research Physicist - Dr. Jen Shu Hsieh

3. Other personnel:

Senior Reactor Operator - MAJ Leonard A. Alt Senior Reactor Operator - SFC Stephen Holmes Ser.ior Reactor Operator Candidates:

CPT Philip Mattson Mr. Robert George (effective 26 Sep 1988)

Mr. Boris Stallings (effective 22 Dec 1988)

4. Departures during CY 1988:

Capt Kenneth Hodgdon (SRO License terminated 1 Sep 88)

AFRRI TRIGA Reactor Facility Page 5 1

j q

- . _ _ - - -- - - - - - . - - - - - - - d

1988 A.W. A L H E H;h T I 5. There were changes to the RRFSC during the 1988 calendar year. D r.

I Naresh Chawla was replaced (26 Apr 1988) by Mr. Thomas O'Brien, as Acting Chairman of the Safety and Health Department (SHD), who later was replaced (5 Dec 1988) by Mr. Douglas Ashby when he assumed the chairmanship of the Safety and Health Department (SHD) during the calendar year. Dr. Frank I Munno resigned (5 Oct 1988) from the committee and was replaced by Dr.

Marcus Voth, Pennsylvania State University. Mr. John Misner, ORI Incorporated, was appointed (18 Dec 1987) as a Special Member to study the software for the new digital reactor console.

The new charter for the RRFSC is enclosed as Attachment A to this re por t.

The 1988 RRFSC consisted of the following membership to satisfy the Reactor Technical Specifications:

I Chairman - CAPT Richard I. Walker (former Deputy Director, AFRRI)

I Regular Members Mr. Douglas Ashby (Chairman, Safety and Health Dept., AFRRI)

Mr. Mark Moore (Chairman, Radiation Sources Dept. and Reactor Facility Director, AFRRI)

I Dr. Marcus Voth (Director, Pennsylvania State University Breazeale Reactor and Professor of Nuclear Engineering, Pennsylvania State University)

I (appointment effective 5 Oct 1988)

Mr. Jathan W. Stone (Head, Safety Directorate, Naval Research Labs)

MAJ David P. Alberth (Radiation Safety Officer, Uniformed I Services University of the Health Sciences) l 3 Special Members g CDR Gary H. Zeman (Chairman, MRA, AFRRI)

Mr. John Misner (ORI Incorporated)

I Observer Mr. John Menke (EPA, Montgomery County, MD)

I Meetings of the RRFSC were held on:

26 Apr 1988 I 26 May 1988 19 Oct 1988 15 Dec 1988 j

1 l

1 I ,

AFRRI TRIGA Reactor Facility Page 6 )

I i

I l 1

v.

L 1988 ANNU AL HEPORT l

Section I Changes to the facility design, performance characteristics, operating

[; procedures, and results of surveillance testing.

l A. DESIGN CHANGES .

A number of facility design changes were completed during CY 1988.

'These_ changes are discussed in more detail in Section V and Attachment C. to this report.

! B. PERFORMANCE CHARACTERISTICS There were no changes in the performance characteristics during the calender year.

l . C. OPERATIONAL PROCEDURES During the year, numerous changes were made to the Operational Proced ures. A complete set of the revised operational procedures as of

'15 Dec 1988 is at Attachment B to this report. Through the brief summary of changes listed 'below, the 1988 operational procedural evolutions can be traced.

Change 1 L Procedure VIII - Tab B: raily Operational Start-up Checklist Procedure VIII - Tab B1: Daily Safety Chechlist (RRFSC Meeting - 26 April 1988)

Change:

1. Remove Item I.7 " Gas Stack Monitor (SGM) & Cooling Air Blower Off". The old SGM was replaced by the new SGM and this air blower was no longer being used.

2. Reword Item VI.7.c "" Gas Stack Monitor High Alarm set to .."

Changed from "1.5 E3" to "800 MPC Ar-41" Based on calibration factors for the SGM system, the instrument setpoint changed for each calibration. However, the 800 MPC Ar-41 requirement does not change. This change provides procedural consistency.

Change 2 Procedure VIII - Tab I: Daily Operational Shutdown Checklist (RRFSC Meeting - 26 April 1988)

Change:

Remove Item VI.10 " Coffee Pot Off" The coffee pot is no longer in the reactor area.

AFRRI TRIGA Reactor Facility Page 7

198P ANNUAL REPORT I

L  !

Change 3 )

D Procedure VI - Emergency Procedures  ;

(RRFSC Meeting - 26 April 1988)  ;

. Change:  !

Reword Item 1.c j

': "EAS Commander" to ERT Commander" Changed for consistency with the new Emergency Plan. I Change 4 Procedure VIII - Tab F: Square Wave Operation (Mode II) l (RRFSC Meeting - 2G April 1988)

Change:

Section 2.b has been elaborated upon to elucidate the procedure required to perform a Square Wave Operation when the TRANS rod is needed to achieve initial criticality.

Change 5 .i Procedure I - Tab A: Reactor Exposure Room Entry Procedure (RRFSC Meeting - 26 April 1988)

Changes:

1. Item 1.c.(2) changed to add the requirement for wrist dosimetry to enter the exposure rooms.

2. Change Item 5.d to delete the phrase "and re-secure yellow area, if necessary". A yellow area painted on the floor was-no longer being used as a restricted area because the entire Prep Area was now secured.

Change 6 Procedure VIII - Tab G: Pulse Operation (Mode III)

(RRFSC Meeting - 19 Oct 1988)

Change:

Item 12 Removed: " Select proper pulse detector according to table below". This was a redundant line.

.AFRRI TRIGA Reactor Facility Page 8

1988 ANNU.\L REPORT I Change 7 I

Procedure VI - Emergency Procedures 1 (RRFSC Meeting - 19 Oct 1988)

Change:

Item 2.b reworded to state " Secure any exposure facility in use so that personnel access to that facility is not possible".

This allowed the reactor staff to adequately secure the facility and evacuate the building more rapidly.

Change 8 I Procedure 0 - Procedure Changes (RRFSC Meeting - 15 Dec 1988)

I change:

Paragraph 5 added "If the entire book of procedures is reviewed, a single signature block on a title page will substitute for individual review".

Change 9 Procedure I - Conduct of Experiments I (RRFSC Meeting - 15 Dec 1988)

Changes:

I 1. Added to Item 1.c the following phr se "and the core position of the experiment facility to be util 1ed". This codified in the procedure a requirement that was already being done.

2. Changed Item 1.d from "or his designee" to " acting RFD or I ROS". This change was made for clarification.

3. Changed department title in Item 1.e to " Military Requirements

& Applications, Operational Dosimetry Division (MRAD)"

I This wat done to correct a misnomer.

I I

I

'"""'""^"""""""'"'" """" "

I I

P 1988 ANNU AL REPORT s

h 1

Chansre 10 1 J- Frocedure I - Tab A - Renctor Exposure Room Entry Procedure  !

l - (RRFSC- Meeting - 15 Dec' 1988)

Changes:

1. Deleted from first 11

• of Item 2.b the. phrase ."and Head, Safety and Health Department or his representative . The .j RFD alone is responsible for granting access to these areas.

L 2. Chboged department title in Item 2.b from " Head SHD" to "Chsirman, Safety and Health Department (SHD)". This was

. done to correct a nisnomer.

3. The remaining correction to Item 2.b clarified on Prep Area .

{

entry status, required personnel listings, ad special entry j conlitions. 1

4. Item 2.c.(2)' was clarified to mean "AFRRI TLD Whole Body f badge". I

5. Item 2.c.(5) was clarified to include " data obtained" as i part of the information available before ER entry. ]

6.: Item 3.b was modified to address exposure room openings in a {

more general manner, but without decreasing safety concerns. 1

7. Additional entry survey. areas were added to Item 3.c.

8. Item 3.c was clarified to address accessible areas and extended stay time for personnel entering the exposure rooms.

9. -Item 5.b was modified to address the case of a non-monitored opening.

10. Item 6.e was reworded to make a technically correct statement.

11. Item 7.c was entirely rewritten to more adequately detail the

. procedure.

12. Item 8.a was expanded to include conditions when the warning horn in the exposure rooms is disconnected.

i d

1 I

1 i

AFRRI TRIGA Reactor Facility Page 10 I

l j

mv-~- -- .__

f. .

1988 \NN:' AI. R El' ORT l

I I

Change 11 f

Procedure I - Tab B: Core Experiment Tube (CET) l (RRFSC Meeting - ;5 Dec 1988)

Changes: '

l

1. 1. tem 1.1 and Item 4.i were modified to ensure that the reactor core pegboard is kept current,

2. Item 2.e and Item 4.c were modified to ensure that the reactor k operator is aware of reactivity changes.

3. Item 3.a clarifies the personnel dosimetry requirements during retrieval of samples from the core.

)

4. The procedure for sample withdrawal from the CET with respect to radiation levels was clarified in Item 3.i.

Change 12 Procedure I - Tab E: In-Pool /In-Core Experiments (RRFSC Meeting - 15 Dec 1988)

Change:

The following line was added " Ensure that a member of the reac tce staff and a SHD representative are present during the removal of samples from in-pool and in-core locations".

This insured Health Physics monitoring of samI;1e removal.

Change 13 Procedure V - Physical Security (RRFSC Meeting - 15 Dec 1988) {

i Change:

The requirement to lock the reactor room during prolonged absences was udded.

Change 14 Procedure VI - Emergency Procedures (RRFSC Meeting - 15 Dec 1988)

Change:

Item 2.b reworded to state " Secure any exposure facility which are in use so that personnel access to that facility is not possible". This allowed the reactor staff to adequately secure the facility and evacuate the building more rapidly.

AFRRI TRIGA Reactor Facility Page 11

-- -. __ _= ._ __ _ . . - _ _ --

L .,

1988 ANNU AL REPORT Ll k.:

' Change 15 L Procedure VIII - Reactor. Operations (RRFSC Meeting - 15 Dec 1988) g~ Changes:

f, 1. Item 2 : now provides- the rationale for performing a Daily {'

- Safety Checklist.

2. Item 3 requires that' the SRO-on-call be ' annotated at the top L each. page in the operations logbook.

3. Item 7 requL es that Daily Operational Shutdown' checklist be performed c =. .he end of each day in which a' Daily Operational l.

Startup Checklist or a Daily Safety Checklist has been R performed.

Change 16 Procedure VIII - Tab A: Logbook Entry Checklist change:

The.following line was added to Item 4.a: "The operator in charge will be designated in the logbook whenever multiple

~

operators are signed on the. console".

Change 17 Procedure VIII - Tab I: Daily Operational Shutdown Checklist Change:

Items VI.7 and VI.9 were reworded to be grammically similar to L

other items in the same section. No performance standards were changed.

D. SURVEILLANCE TESTING All surveillance items were accomplished on time. Malfunctions discovered during operations are discussed in 'Section IV.

l l

J AFRRI TRIGA Reactor Facility Page 12 1

- _ - - - _ - _ _ _ _ _ _ _ = _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ - _ _ . .

b 1988 ANNU AL REPORT

),

t ..

)

Section II

[

L Energy generated by current reactor core f'

MONTH kwhr JAN 2756.4 l FEB. 6610.1 MAR 2294.6 APR 1665.7 MAY 2362.7 l

' JUN.- 5620.2 JUL 3030.2

-AUG 2093.9 SEP 2112.4 OCT 1433.5 NOV 1736.9 DEC 255.1 Total 31971.7 _

Total' energy generated ~ this year 31971.7 kwhr Total energy on thie core 629716.2 kwhr Total pulses this year > $2.00 211 Total pulses on this core > $2.00 4093 Section III Unscheduled shutdowns There were no unscheduled shutdowns during this reporting period.

I AFRRI TRIGA Reactor Facility Page 13 4

n

I 1988 ANNU AL' REPORT f-l .Section IV Safety related corrective maintenance

l. .

L The following are excerpts from the malfunction logbook during the reporting period. The reason for the corrective action taken, in all cases, was to return the failed unit to its proper operational status.

7 Jan 88- Problem: While securing the ' Core Experiment Tube into 3

storage position S-6, the bulk water temperature connecting wire was knocked loose.

Solution: The wire was repaired. The system was tested and found to be operational.

i-13-14 Jan 88 Problem: The meter of the NV/NVT circuit was indicting a false reading when the left-hand cor. sole drawer was moved (opened / closed).

Solution: A connector to the NV/NVT circuit board was replaced. The NV/NVT system was calibrated / tested and found to be operational.

1-2 ' Fe b 88 Problem: The Square Wave mode of operation was found to be non-operational during testing.

JSolution: Investigation showed that a blown fuse was the problem. The fuse was replaced. The system was tested and found to be operational.

4-5 Feb 88 Problenl: Air was slowly leaking from the primary compressor that supplied compressed air to the Transient rod air system.

Solution: A loose bearing was found to be the cause of the leaking air. While this bearing was replaced, the air supply was temporarily switched to ,* s back-u p compressor system. After the bearink was replaced and tested, the air supply was returned to ' he primary compressor. The air system was testen ,er leakage and was found to be functioning properly. i AFRRI TRIGA Reactor Facility Page 14 l

i l

19R8S A>N!!AL REPOR T 6- May 88 Problem: During a routine pulsing operation, the console indicated that the Transient rod did not SCRAM after the

[ pulse. The console also indicated that the remaining .

rods had SCRAMMED, leaving the reactor in a suberitical condition.

I Solution: A preliminary investigation showed ' that the -

Transient rod did SCRAM following the pulse operation.

7

' The SCRAM function was fully operational. However.

further investigation showed that the wiper arm of.the rod down microswitch was broken. The microswitch was replaced. The Transient rod system was tested and

(~ found to be operational.

2 Aug .88 Problem: The low level alarm light on the Stack Gas Monitor .became illuminated. Although no operations were currently being conducted, further operations were administratively prohibited. ,

Solution: The circuit boards were removed and re-seated. The system was reinitialized, tested, and found to be operational. -

10 Aug 88 Problem: During the Daily Operational Start-up, Hemote Area Monitor E-3 was found to be non-operable. Proper exposure levels were indicated but the monitor failed in the alarm test position.

Solution: RAM E-3 was replaced by another' calibrated

. RAM. Prior to any operations or openings of Exposure Room 1, the system was tested and found to be operational.

l

}

1 f1 AFRRI TRIGA Reactor Facility page 15 i

l L

L

f' 1988 ANNU AL REPOf<T j l

l l

1 29-30 Nov 88 Problem: During monthly maintenance checks, the air '

)- dampers in the ventilation system failed to function L no: mally, i.e., the dampers did not open.

Soluilon: Upon investigation, a solenoid that supplied i- air to open the dampers was r.ot operational. Air pressure to the solenoid was checked. The air lines to the dampers were bled and the damper mechanisms were

' lubricated. Upon the reapplication of air, the system was tested and found to be operational.

l 1 Dec 88 Problem: The Stack Gas Monitor (SGM) was turned off and for replacement of the cpu board with a newer version 5 Dec 88 supplied by the manufacturer. The CPU board was replaced. A QA test was performed and the system was found to be operational. However, the printer printed constantly without stopping.

S_olu tion: Notified the manufacturing representative in cr.ler to fix the printing problem. They stated that this was a minor software problem which in no way affected the monitoring capability or operability of the SGM. The programmable ROM chips were replaced. The printer was tested and functioned properly. The SGM was reinitialized, tested and found to be operational.

AFRRI TRIGA Reactor Facility Page 16 )

I l

1

1968 BNI.AL k ElNI< 1 I Section V Facility change: aanges to procedures, and new experiments A. The 10 CFR 50.59 safety reviews of; the new reactor instrumentation and control system, the warning lights in the control room from the Primary Continuous Air Monitor and the Stack Gas Monitor system and the Cerenkov I detector, and t9 digital voltmeter are included as in Attachment C to this re po r t.

I B. A safety review was also performed concerning the relocation of the Equipment Room 3152 roof hatch, as shown in Attnehment C to this report.

Movement of the roof hatch served to improve the drainage on the reactor J

1 roof. In addition, the main door entrance to the reactor facility is being I moved outward so that the door to Equipment Room 3152 falls inside the main door entrance to the reactor facility. Also, the ceiling above the relocated doorway is lined with wire mesh to further enhance the reactor physical security. The work on the relocation of the main door entrance to the reactor facility is still in progrese and is expected to be completed by the end of  ;

1989. The safety reviews performed on these two facility modifications showed j that no unreviewed safety question existed. <

C. A complete set of revised Operating Procedures is included as Attachm ot B to this report.

D. The new experiment performed during CY 1988 is covered in Attachment E.

Section VI Summary of safety evaluation changes not submitted to NRC pursuant to 10 CFR 50.59 Attachment C satisfies the requirements of this section. Each modification is described and the basis for the conclusion that each change involves no unreviewed safety question, and that there are no changes to the Technical I Specifications, has been provided.

The additional modification concerning the movement of the Equipment Room 3152 roof hatch, addressed in Section V, is not included in the SAR nor the Technical Specifications. Neither is it required for the safe operation of the reactor. However an analysis was performed and shows that no unreviewed safety questions exists.

License Event Reports are included as Attachment D to this report.

I AFRRI TRIGA Reactor Facility Page 17 I l l

1 I -_

h' 1988' ANNUAL REF ORT 1? <

Sec' tion VII Summary of radioactive effluent released

~A. ' Liquid Waste The reactor has produced no liquid waste during CY.1988.

. B. Gaseous Waste - There were no particulate discharges in CY 1988. The 1 total Argon-41 discharges in CY 1988 was 9.145 Cl.

C. Solid Waste . - All solid material was transferred to the AFRRI byproduct -

license: none was disposed under the R-84 license.

h ' Section VIII-Environmental radiological surveys A. . Th' e environmental sampling of soil, water, and plant growth reported radionuclides levels that were not demonstrable above the normal range. The radionucildes that were detected were those normally expected from natural background and from long-term fallout.

B. The environmental monitoring (dosimetry) program reported the following results for CY 1988:

1. The average background of about 20 Thermoluminescent dosimeters .

(TLD) located within a 15 mile radius from the AFRRI site was determined to

.be 99.49 +/- 3.80 ' millirem. -

2. The average reading of approximately. 30 environmental stations located on the AFRRI site from background was determined to be (-5.42 +/-

!. 2.05) millirem.

3. The single highest environmental station reading was (15.72 +/- 19.31) millirem above background.

4. ' The above results are expressed at a 95% confidence level.

C.- The in-plant surveys, including analysis of effluent filters, showed no i l

measurable attivity (except as reported in Section VII) in all areas outside the normal restricted-access areas.

D. There were no special environmental studies conducted during this year.

l l

7

.AFRRI TRIGA Reactor Facility Page 18 i

l~ j l

1988-ANNIAL REPORT l l

l l

I Section IX Exposures greater than 25% of 10 CFR 20 limits j

1 i

There were no exposures to staff or visitors greater than 25% of 10 CFR 20 limits. )

g.

I l

.I I

I I

I

I I

I I

I I

AFRRI TRIGA Reactor Facility Page 19 I

I

,___.-._,_-__,___,-___-_________,._7__-____. _ ______-__, __.,._,

i-

[.

ATTACHMENT A CHARTER FOR THE ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE

, REACTOR AND RADIATION SAFETY COMMITTEE l

)

I-

).

1 r

i

i DEFENSE NUCLEAR AGENCY ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE BETHESDA, MARYLAND 20814 5145

)

i AFRRI/DIR' 3 October 1988 t

SUBJECT:

RRFSC Charter l

TO: RRFSC Committee

1.

Reference:

Ammendent Charter

2. This Charter is now in affect until further notification.

1 bj GEORGE . IRVING, III l Colon USAF. BSC Direc r

0 @* 7 yc. ,.

l CHARTER FOR THE ARMED FORCES RADIOBIOLOGY R AICE YN hTE T REACTOR AND RADIATION FACILITY SAFETY COMMITTEE hg) et 190 se flRFU I. INTRODUCTION: p f This Charter governs the Armed Forces Radiobiology Research Institute's (AFRRI's) >

Reactor and Radiation Facility Safety Committee (RRFSC). The charter codifies the requirements of the AFRRI TRIGA Reactor Technical Specifications, complies l with descriptions in the Safety Analysis Report for AFRRI TRIGA Mark-F Reactor, and adheres to the guidelines contained in American National Standard Institute (ANSI)/ American Nuclear Society (ANS) Standard 15.1-1982. In the execution of the duties and functions premibed by Department of Defense Directive number

) 5105.33 (

Subject:

Armed For n Radiobiology Research Institute, dated 25 NOV 87),

the AFRRI Director is the pyvinent for this charter and is the sole authority for changes to or deviations from this document. I 1

II. PURPOSE and AUTHORITY:

The RRFSC is directly responsible to the AFRRI Director. The committee reviews items that could affect the radiological health and safety aspects of facility operations and makes recommendations to the AFRRI Director concerning the i following sources at the Institute:

TRIGA Reactor LINEAR Accelerator Cobalt Facility Theratron Facility X-Ray Facility Other radiation sources as designated by the AFRRI Director.

The committee's review oversight includes: The physical facilities, the planned operations, and the qualifications cf supervisory and operating personnel which relate to the safety of the Institute, its staff, the public, and the environment. The RRFSC discharges its principal responsibility of broad ovenight of the Institute's radiation sources health and safety issues through the review of studies and reports prepared by the staff of the Radiation Sources Department and Safety and Health Department, and by commissioning periodic external audits of key operations.

Additionally, the RRFSC advises the Reactor Facility Director, the Radiation Sources Department Chairman, and the Safety and Health Department Chairman in those functional areas specified in section V of this document.

IIL COMPOSITION OF COMMITTEE, QUALIFICATIONS AND TERMS OF SERVICE:

A. Committee Chairman, as appointed from the AFRRI Directorate by the AFRRI Director. Term as appointed by AFRRI Director.

B. Reactor Facility Director. Permanent with position.

C. Radiation Sources Department Chairman (if different from Reactor Facility

l. Director). Permanent with position.

D. Safety and Health Department Chairman. Permanent with position.

E. One to three non-AFRRI members, appointed by the AFRRI Director, who are knowledgeable in fields related to reactor safety. At least one shall be 1

a Reactor Operations Specialist, or a Health Physics Specialist. Annual l term, renewable by AFRRI Director's appointr9ent.

F. Special RRFSC Members (Temporary Members):

1. Other knowledgeable persons. to serve as alternates for those in paragraph III.D., as appointed by the AFRRI Director.

Temporary, for duration specified by AFRRI Director.

2. Voting ad hoc members, invited by the AFRRI Director, to assist in review of a particular problem. Temporary, for duration specified by AFRRI Director.

3. Non-voting members as invited by the RRFSC Chairman.

Temporary, as indicated by RRFSC Chairman.

G. The minimum qualifications for a person on the RRFSC shall be 6 years of professional experience in the discipline or specific field represented. A baccalaureate degree may fulfill 4 years of experience.

IV. MEETINGS and RULES:

A. Special Members (Alternates)

Alternate members may be appointed in writing by the RRFSC Chairman to serve on a temporary basis. No more than two alternates shall participate on a voting basis in RRFSC activities at any one time.

B. Meeting Frequency The RRFSC, or a subcommittee thereof, shall meet at least four times a i calendar year. The full RRFSC shall meet at least semi-annually.

C. Quorum A quorum of the RRFSC for review shall consist of the Chairman or designated alternate) and two other members (or alternate members , one of i

which must be a non-AFRRI member. A maaority of those present shall be regular members.

I D. Voting Rules Each regular RRFSC member shall have one vote. Each special appointed 7

member shall have one vote. The majority is 51% or more of the regular and special members present and voting.

a l

1 I

E. Minutes l

Minutes of the previous meeting shall be available to regular members at least I week before a regular scheduled meeting. i F. Subcommittee An RRFSC Subcommittee will consist of the regular RRFSC Chairman (or  !

f his designated substitute), one other permanent member, and selected I special members as designated by the regular RRFSC Chairman. No more than fifty percent of the voting attendees may be special (temporary) i g memoers or alternate members. I V. FUNCTIONS:

A. REVIEW OF REACTOR OPERATIONS I The RRFSC shall review the items specified at sub-paragraphs 1 through 10 below, and will concur or non-concur with the fmdings of the Reactor Facility Director that the items present no hazard to the public health and safety. The review will normally occur at the first meeting following the I implementation of actions for which the Reactor Facility Director has determined that no unreviewed safety questions exist, and prior to implementation of actbus for which the Reactor Facility Director has g determined that an ureviewed safety ouestion exists.

A written report or minutes of the fmdings and recommendations will be I submitted to the AFRRI Director in a timely manner after the review has been completed.

1. Safety evaluations for (1) changes to procedures, equipment, or systems and (2) tests or experiments conducted within NRC approval under provisions of Section 50.59 of 10 CFR Part 50, to verify that such actions did not constitute an unreviewed sdety question.

2. Changes to procedures, equipment, or systems that change the original intent or use, and are non-conservative, or those that involve an I unreviewed safety question as defined in Section 50.59 of 10 CFR Part 50.

3. Additions and modification to, and testing procedures for, SAR stated I systems including the ventilation system, the core and its associated support structure, the pool, coolant system, the rod drive mechanism, or the reactor safety system; unless the additions and modifications I are made and tested to the specifications to which the systems were originally de:igned and fabricated.

4. New experiments of a type for which previous authorization has not been granted, for radiological safety, prior to issuance of a reactor authorization.

5. Proposed tests or experiments that are significantly different from previously approved tests or experiments, or those that might involve l

(

a i

I l

an unreviewed safety question as defined in Section 50.59 of 10 CFR Part 50.

6. Proposed changes in technical specifications, the Safety Analysis Report, or other license conditions.

7. Violations of applicable statutes, codes, regulations, orders, technical specifications, license requirements, or of internal procedures or

{

instructions having nuclear safety significance.

8. Significant variations from normal and expected performance of facility

l. equipment that might affect nuclear safety.

9. Events that have been report *d to the NRC.

10. Audit reports of the reactor facility operations.

B. AUDIT OF REACTOR OPERATIONS Audits of reactor facility activities shall be performed under the cognizance of the RRFSC, but in no case by the personnel responsible for the item audited, annually not to exceed 15 months. A report of the findings and recommendations resulting from the audit shall be submitted to the AFRRI Director within three months after the audit has been completed. Audits may be performed by one individual who need not be an RRFSC member. These audits shall -mine the operating records and the conduct of operations, and shall encompass the following:

1. Conformance of facility operation to the Technical Specifications and the license.

2. Performance, training, and qualifications of the reactor facility operations staff.

3. Results of all actions taken to correct deficiencies occurring in facility equipment, structures, systems, or methods of operations that affect safety.

4. Facility emergency plan and implementing procedures.

5. Facility security plan and implementing procedures.

6. Any other area of Facility operations considered appropriate by the RRFSC or the AFRRI Director.

7. Reactor Facility ALARA Program. This program may be a section of the total AFRRI program.

C. REVIEW OF OTHER RADIATION FACILITIES OPERATIONS:

The RRFSC shall review the items specified at sub-paragraphs 1 through 7 below and will concur or non-concur with the findings of the Radiation Sources Department Chairman that the items present no hazard to the public health and safety. The review will normally occur at the first meeting following the r

l I

implementation of actions for which the Radiation Sc c.es Department Chairman  !

has determined that no unreviewed safety questions ex :, and prior to

) implementation of actions for which 'the Radiation Sources Department Chairman {

has determined that an unreviewed safety question exists.

A written report or minutes of the findings and recommendations will be submitted to the AFRRI Director in a timely manner after the review has been completed.

I

1. Changes to routine authorized irradiations or procedures with respect l to personnel safety for the linear accelerator, cobalt-60 facility, therstron, and other non-reactor major radiation sources.

2. Qualifications of new source operators, for compliance with appropriate regulations and guidelines governing training for operations of a

! particular source.

3. Proposed amendments to any USNRC Licenses governing the use of the other sources.

l

4. Violations of application statutes, codes, regulations, orders, license requirements, or of internal procedures or instructions having nuclear safety significance.

5. Significant variations from normal and expected performance of facility equipment that might affect nuclear safety.

6. Events that have been reported to the NRC.

7. Audit reports of facility operations.

VI.

REFERENCES:

a. Technical Specifications for the AFRRI Reactor Facility, Docket 50-170, License R-84, June 1984.

b. Safety Analysis Report for AFRRI TRIGA Mark-F Reactor, June 1987.

c. American National Standards Institute /American Nuclear Society Standard 15.1-1982, The Development of Technical Specifications for Research Reactors, September 1982.

s

I I

I I .11 cameNr e CURRENT I REACTOR OPERATING PROCEDURES I

I

~

I l

I I

I I

l I 1 i

I

f::

I '

REACTOR OPERATING PROCEDURES AFRRI TRIGA MARK F REACTOR L

1 APPROVED BY THE REACTOR FACILITY DIRECTOR:

OfiiGliiA1. SiGhpiD MARK' MOORE Date 1

l 1

l l

'l

'l

.l l

e l 1

b

'I l

i

! I k Revised: 15 Dec 1988 )

i 1

1 REACTOR OPERATING PROCEDURES l

INDEX -

l 9

0. PROCEDURE CHANGES .............................. 15 Dec 1988  !

I I. CONDUCT OF EXPERIMENTS ......................... 15 Dec 1988 TAB A - Exposure Room Entry .................... 15 Dec 1988 TAB B - Core Experiment Tube (CET) ............. 15 Dec 1988 TAB C - Extractor System .......................... Dec 1987 TAB D - Pneumatic Transfer System (PTS) ........... Jan 1984 TAB E - In-Pool /In Core Experiments ............ 15 Dec 1988 II. REACTOR STAFF TRAINING ............................ Jan 1985 III. MAINTENANCE PROCEDURES ............................ Jan 1985 IV. PERSONNEL RADIATION PROTECTION .................... Jul 1982 V. PHYSICAL SECURITY .............................. 15 Dec 1988 VI. EMERGENCY PROCEDURES ........................... 15 Dec 1988 l VII. REACTOR CORE LOADING AND UNLOADING ........ ... ... Dec 1987 VIII. REACTOR OPERATIONS ............................. 15 Dec 1988 TAB A - Logbook Entry Checklist ................ 15 Dec 1988 TAB B - Daily Operational Start-up Checklist ...... Apr 1988 TAB B1 - Daily Safety Checklist ...... ............ Apr 1988 TAB C - Nuclear Instrumentation Set Points ........ Apr 1988 TAB D - K-Excess .................................. Jul 1982 TAB E - Steady State Operation'(Mode I/IA) ........ Jan 1985 TAB F - Square Wave Operation (Mode II) ........... Apr 1988 TAB G - Pulse Operation (Mode III) ................ May 1988 TAG H - Weekly Operational Instrument Checklist ... Oct 1984 ,

TAB I - Daily Operational Shut-down Checklist .. 15 Dec 1988 l TAB J - Reactor Monthly Usage Summary ............ .Jul 1982 ,

TAB K - Stack Gas Monitor Procedure ............... Dec 1987 l 1

IX. REACTOR ROOM SAFETY ............................... Dec 1986 ]

1 l

l

- Revised: .15 Dec 1988 PROCEDURE O PROCEDURE CHANGES l- General: This establishes procedures for. permanently or temporarily changing l l reactor operating' procedures.

Specific:

- 1. Permanent changes are made by revising the' entire procedure. .The revised procedures will be approved by the Reactor Facility Director (RFD)

. and reviewed, by the Reactor and Radiation Facility Safety Committee (RRFSC).

2. Temporary changes may be made in ' pen and ink on the current

~

procedure when initialed by the RFD or Reactor Operations Supervisor (ROS).

These' changes must be documented and subsequently reviewed by the RRFSC at the next scheduled meeting. ,

3. Temporary procedures may be established by the RFD for a specific.

situation.

4. All procedures -(temporary or permanent) will have an initial block for all  ;

operators and reactor staff members. When the initial block is completed, the

- procedure will be placed in the Reactor Operation Binder and kept available for operator review.

5. If the entire book of procedures is reviewed, a single signature ' block .on a title page will substitute for individual review.

I e I Revised: 15 Dec 1988 PROCEDURE I CONDUCT OF EXPERIMENTS General:

1. All experiments will be observed during irradiation with the exception of CET experiments or those in which no movement is possible. The closed-circuit televisions (CCTV's) in the exposure rooms and over the reactor pool can be used to meet this requirement.

2. All experiments will be set up so as to preclude movement unless the experiment apparatus is designed for movement (such as rotators, etc.).

I 3. The Reactor Staff will conduct a thorough inspection of all experiments to determine that no unauthorized materials, items or substances, or equipment are irradiated.

4. ALARA will be practiced during all experiments.

Specific:

1. Experiment Review (processing of Reactor Use Request (RUR)):

a. Check RUR for completeness (Section I should be filled out),

b. Check experiment protocol against reactor authorizations. Assign Ieactor authorization number.

c. Fill-in Section II of RUR with special instructions, as app'ropriate.

Assign RUR sequence number. Write in estimated or measured experiment worth and the core position of the experiment facility to be utilized in the I. appropriate block (lower lef t-hand corner of form).

d. Have the Reactor Facility Director (RFD), acting RFD, or Reactor g Operations Supervisor (ROS) review and sign the form.

3 e. Forward the RUR to the Military Requirements & Applications Department, Operational Dosimetry Division (MRAD) and the Safety & Health Department (SHD) for coordination.

f. Ensure the RUR form is returned prior to irradiation.

2. Conduct of Experiments. Perform setup and irradiation of experiments in accordance with the following procedures:

I a. Exposure Room Entry - TAE A.

b. Core Experiment Tube (CET) - TAB B.

c. Extractor System - TAB C.

I d.

e.

Pneumatic Transfer System (PTS) - TAB D.

In-pool /In-core Experiments - TAB E.

3. Complete the RUR by filling out Section IV with the appropriate information.

4. Attach form to clipboard in the control room.

I I

I I .

'ie B'

Revised: 15 Dec 1988 I TAB A: REACTOR EXPOSURE ROOM ENTRY PROCEDURE

1. REFERENCES

a. 10 CFR 20, " Standards for Protection Against Radiation"

b. USNRC licenses: R-84, 19-00330-02 I c. AFRRI Radiological Safety Instructions

2. GENERAL

a. PURPOSE: This procedure specifies all safety and security procedures for activities involving entry into the AFRRI TRIGA Reactor exposure rooms, currently designated exposure rooms 1 and 2 (rooms 1123 and 1122).

b. AUTHORIZED ENTRY: Both green and orange badged personnel, may enter a reactor exposure room under the supervision of the Reactor Facility I Director (RFD) or his representative. Visiting personnel (V badge) require special authorization by both the Chairman, Safety and Health Department (SHD) and RFD to enter either exposure room. In general, permission to enter I the exposure rooms will be granted personnel whose duties require such entry, however permission may be denied to personnel for serious or repeated safety or security violations, or for safety reasons emanating from conditions in the exposure rooms themselves. All personnel who are granted either I escorted or unescorted access to the prep area or warm storage will receive a special prep area safety briefing prior to being granted access. Only personnel who have been granted unescorted access will be given the combination to the prep area or warm storage. The RFD is responsible for maintaining two seperate rosters in the prep area: one roster for personnel who have been granted unescorted access, and one roster for personnel who I have been granted escorted access. Other personnel requiring unescorted access to the prep area or warm storage for a specific purpose or time period may be granted special access in writing by the RFD with concurrence of SHD.

However, these personnel who are granted specical access from the RFD will not be given the combination to the prep area.

c. ER ENTRY INSTRUCTIONS - All personnel will:

(1) Know the Reactor staff representative is in charge of all I operations in the prep area. Obtain permission to enter either exposure room from the Reactor staff representative.

(2) Wear AFRRI TLD whole body badge, wrist dosimeter, and pocket  ;

I dosimeter.

(3) Wear booties, eye protection, gloves and coat.

(4) Check and log pocket dosimeter reading on log in prep area prior I to entry.

(5) Familiarize themselves with approximate radiation levels in the room, based on radiological surveys performed and data obtained by SHD.

(6) Ensure that all materials removed from the exposure room are I properly labeled and entered on the exposure room entry log AFRRI FORM 130 (enclosure 2 of this procedure), and the activated materials control log.

(7) Glove and coat requirements may be waived, by the Reactor I

I .

P i

)

l Representative on an individual basis, for personnel who will not be touching anything in the exposure room. There must be a specific reason for waiving l such requirements)

d. DEPARTURE FROM REACTOR EXPOSURE ROOM ENTRY PROCEDURES: Any departure from the following procedures will require a special work permit (SWP). Exceeding any radiation dose limits will require a written justification

)

from the supervisor of the research project which must be approved by the Head, SHD.

l l

3. SHD EXPOSURE ROOM SURVEY

a. EXPOSURE ROOM CAM: Prior to opening either exposure room, the respective CAM must read 2000 cpm or less, above background. If the CAM reads 2000 cpm or greater above background, change the filter of the CAM.

If 10 minutes or more have lapsed since the end of the reactor run, the door may be opened to the first step to facilitate radioeffluent clearance in the room. Then check the CAM after 1 minute and if the reading is below 2000 cpm above background, proceed with the exposure room opening. If its above, change the filter and wait another minute. If the CAM alarms during or immediately after a run, change the filter and reset the CAM.

b. DOSE RATE AT FACE OF DOOR: If the dose rate at the face of the plug door in the direct line of sight of the reactor tank bulge reads greater than 100 mr/hr, the door will be closed sufficiently to preclude access. The plug door will be reopened upon agreement of the SHD and RFD representatives for reevaluation of radiation levels.

c. DOSE LEVELS IN ROOM: Exposure rates will be measured at specific sites in the rooms. These measurements will be given to both the reactor representative and the personnel entering the room. Additionally the readings will be entered in the room entrance log ( AFRRI FORM 130) and kept in the prep area. The levels will be measured at:

(1) The reactor door face in the direct line of sight of the reactor tank bulge (2) At the contamination line in the entrance of the room (3) The middle of the room (4) One meter from the tank wall or shield (5) Contact with the tank wall or shield (6) The area (s) where individual (s) will be working for an extended period of time and any other place deemed necessary by the SHD or reactor rep re sentatives,

d. ROUTINE ENTRY: Entry is routinely permitted only when the maximum  !

reading in any occupiable area is 1 R/h or less. Entry may be permitted if j levels are 1-5 R/h, but no work will be permitted in fields over 1 R/h. When ]

working in a specific area for any extended time is expected the dose rate in i that area will also be measured and recorded. I (1) If any accessible area inside the exposure room reads over 100 l mR/hr (closed window), where extended stay time is possible, the SHD monitor will remain in the prep area until the room is closed. All personnel entering will be assigned a stay time if they will be working in the high radiation area. l AFRRI limits of 100 mR/ week anti 50 mR/ day are to be used as the basis of j l

I

s stay time determinations.

(2) All exposure room entries will be checked by the SHD monitor for compliance with radiation safety aspects of applicable Reactor Use Requests

( RU R's). If not, non-compliance will be reported to RFD and to SHD.

e. FILLING OUT THE SURVEY OF EXFOSURE ROOM OPENING LOG: The exposure room opening log sheet must be filled out completely for each I opening of an exposure room (see enclosure 2).

each blank on the entry log sheet, if a section is not applicable to the Care must be taken to fill out particular opening, N/A should be filled in the blank.

4. NON MONITORED OPENING:

I a. The exposure rooms may be opened without a SHD monitor present if ALL the following conditions hold:

(1) The reactor has not been to power in that ER since the last I survey.

(2) The last survey indicated that there were no radiation levels in excess of 100 mR/hr in any area of the ER where extended stay time is possible.

(3) Survey meter readings at the door indicate safe entry conditions (should be less than 1 mR/hr).

(4) The ER CAM should be observed, and its reading (net) should be less than 200 cpm above background,

b. An entry will be made in the exposure room log by a reactor staff member, with a note that the survey has been waived.

c. SHD must be notified if any radioactive materials or equipment are to be removed from the prep area.

5. PERSONNEL PROTECTION PROCEDURES

a. Dosimetry and protective clothing requirements are given in paragraph 2.c, entry instructions.

b. Entry is permitted only after the SHD monitor has completed the survey and reported results to those about to enter (excluding non-monitored openings - Reference Paragraph 4, above).

c. All personnel shall read and log dosimeters when leaving the exposure room using the dosimeter log in the prep area. Net doses over 10 mrem must be reported to the SHD Monitor,

d. Protective clothing will be removed in such a way as not to contaminate

" clean" areas by items from " dirty" areas.

e. All personnel entering the prep area will " frisk" themselves before leaving the prep area.

I

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

l-l

6. SPECIFIC ACTIONS TO OpEN EXPOSURE ROOM DOORS l

a. Turn up exposure room lights (this can be waived for experiment j needs).

b. Check plug door tracks for obstructions; insure all obstacles are clear of the door (including ropes).

c. Secure both entrances to the prep area.

d. Insure that only authorized personnel (see 2.b.) are present in the reactor prep area during exposure room openings.

e. When facility safety interlocks and opening procedures have been satisfied, insert key into exposure room door key panel and open door. DO NOT LEAVE KEY IN LOCK UNATTENDED.

f. Open door in accordance with entry procedures. Ensure all required data is logged in entry log.

g. Ensure that individuals that will be moving lead, bismuth, or other heavy materials are wearing steel-toed shoes.

h. Limit exposure times of all personnel entering the exposure rooms based on the results of the radiation survey.

7. ACTIVATED MATERIALS

a. PLACING MATERIAL IN EXPOSURE ROOM: Before placing any equipment or material in an exposure room for irradiation the following will be observed:

(1) Equipment tagged as AFRRI property: a DF must be sent to both the RFD and the AFRRI property officer. The DF must state that the equipment is knowingly being irradiated and therefore request that it be removed from the property books. It must also state that should the material remain byproduct material after a reasonable amount of time it will be disposed of as radioactive waste. The DF must contain all nomenclature as well as an adequate description of the equipment in order for it to be identified on the property book.

(2) Non tagged AFRRI equipment or material (to be returned): a DF or statement on the reactor RUR must be sent to the RFD giving the kinds and amounts of byproduct material expected to be produced (that is the material that the experimenter wishes to be returned) and a copy or number of their radionuclides authorization number. The DF or RUR statement must be specific and contain an accurate description of the material being exposed (converted to byproduct). Other information will be required from personnel before any material is allowed to be removed from the prep or warm storage areas (see next section of this procedure 7.b. and 7.c.)

(3) Non tagged equipment or material (not to be returned): A DF or statement on the RUR that the experimenter understands that byproduct material produced as a result of their irradiations will be disposed of as radioactive waste, and additionally any material not specifically requested to be held will be disposed of as radioactive waste in the next shipment.

(4) Non AFRRI owned equipment / material: A signed memorandum from

the responsible property owner that they understand that byproduct materials generated in excess of their license will be disposed of as rad waste unless prior arrangements have been made with the reactor /SHD staffs for storage.

I Any material not removed within a reasonable amount of time will automatically be disposed of as radioactive waste.

b. SURVEY OF MATERIALS COMING OUT OF EXPOSURE ROOM (1) All material leaving the exposure rooms must be surveyed for activation or contamination. Survey meter readings will be used to determine I dose levels. Smear surveys may be used, if the SHD representative deems them necessary. All materials will be labeled appropriately in accordance with HPP 0-2 and enclosure 1 of this procedure.

(2) All special equipment that has been activated such as chambers, I rotaters, motors, meters, etc., will be stored under the control of the reactor license or the AFRRI byproduct license in warm storage or the prep area.

Removal of items from the prep area will only be allowed in accordance with the disposition of activated materials, section 7.c. of this procedure.

c. DISPOSITION OF ACTIVATED MATERIALS (1) All materials coming out of the exposure rooms will fall into one of two categories. Category one consusts of materials that are to be removed from the prep area and category two are those materials designated to remain in the prep area. Prep area materials should be tagged with a yellow radiation I material label, or tag filled in " Prep Area Materials" or painted yellow.

Materials labeled or painted yellow are not to be removed from the prep area without SHD approval.

(2) If tagged material must be returned to the exposure room before it has been cleared from the activated materials log, return the material with the label to the prep area before the exposure room opening. At the time of the exposure room opening, give the tag to the SHD representative who will then I clear the materials from the log. When the materials come out of the exposure room a new log entry will be made and a new number assigned the materials.

(3) When materials to be remove from the prep area come out of the exposure rooms, the materials must be tagged appropriately and an entry must be .nade in the activiated materials control log. The tagging procedure and information that must be entered in this log is as follows:

(a) ITEM NUMBER: will be assigned by the SHD monitor in sequence I and be prefixed by an "R" for reactor, "L" for Dr. Ledney "Z" for CDR Zeman, etc. The next character in the item is the calender year. The last character is the sequential item number.

EXAMPLE: L87-005 "L" indicates Dr. Ledney is the Priciple Investigator "87" is the calender year "005" indicates it is the fifth activated item cemoved from the prep arfea by Dr. Ledney in 1987 Each item that has been activated must be assigned an activated item number, J 1.e., if a dog was activiated in a plactic cage, both the dog and the cage must j be numbered and tagged; or if an activated camera is tagged, both the lens and the camera must be numbered if the lens is removed.

1 I

. I

__-__-----_____-__-A

1 NOTE: All labels must be kept with the materials until the materials are disposed of as regular waste or as radioactive waste, or cleared by SHD In any case the tags must be returned to SHD to facilitate removing activated items from tne log.

(b) ITEM DESCRIPTION, AFRRI NUMBER, SERIAL NUMBER, enter a brief description of the item removed. The AFRRI number and serial number shall be entered if applicable.

(c) BACKGROUND LEVELS, enter the background radiation levels of the area where the survey is conducted.

(d) LEVELS ON CONTACT CLOSED WINDOW, LEVELS ON CONTACT OPEN WINDOW, enter the radiation levels detected on the surface of the item being surveyed using both open and closed windows.

(e) SMEAR RESULTS, enter the results of the smear test if it was taken, see section 7.b.l. of this procedure.

(f) LOCATION MATER'/ L REMOVED TO: enter the lab or area to which the materials are being t. 2n. Ensure that the lab is qualified to hold the radioactive materials in accordance with enclosure one to this procedure, all appropriate radiological Safety Instructions and Health Physics Procedures.

(g) PERSON REMOVING MATERIALS, enter the name of the investigator or technician that is taking the materials, ensuring that the person is listed under the principal investigators authorization for handling radioactive materials.

(h) SIGNATURE OF PERSON REMOVING MATERIAL, have person receiving custody of the materials sign the log with the understanding that the materials being received have been activated.

(i) INITIAL OF SHD PERSONNEL, the person that released the activated materials will initial here.

(j) REMOVE FROM LOG, this space is to be checked off when the radioactive material has decayed below activation action levels indicated i..

enclosure 1 to this procedure.

8. COMPLETION OF ENTRY

a. The Reactor Staff Representative will check to see that all personnel have left the exposure room before the plug door is closed. In the event that the warning horn in either exposure room is disconnected, for testing or experiment requirements, the exposure room plug door shall not be closed until at least two (2) licensed reactor operators visually inspect the room to insure thet no personnel remain in the room. To ensure compliance the the reactor Technical Specifications, the names of these licensed operators present at the exposure room closing shall be entered into the reactor operations logbook and on AFRRI FORM 130. At the completion of the test or experiment the warning horn shall be reconnected and tested. All actions regarding the warning horn shall in entered in GREEN ink in the reactor operations logbook.

b. The SHD monitor will not leave the area while the plug door is open without notifying the Reactor Staff Representative.

l

c. Lock the exposure room door control panel; reset lights, if appropriate.

d. Resecure the prep area on departure.

i w k j~-

Revised: 15 Dec 1988 TAB B: CORE EXPERIMENT TUBE (CET)

General: ALARA principles will be practiced during CET operations.

I' Specifics: j

1. CET Insertion into the core: i

a. Ensure a reactor operator is monitoring the reactor console. l

b. Ensure a reactor staff member is present in the reactor room. I

c. Establish communications between the reactor room and the control room.

d. Test fuel-handling tool for operability.

e. Lower the fuel-handling tool into the core and attach to element F28. Notify operator on the console that you are prepared to lift fuel element.

When acknowledged, mt fuel element from the core.

f. Transfer element to a storage rack location and secure fuel-handling tool cable.

I g.

h.

Loosen CET bracket bolts and remove CET bracket.

While the CET is held down, cut cable ties from around the CET.

i. Lift CET from the storage rack location and transfer to the reactor carriage, ensuring that the CET remains as low in the water as possible.

I j. Notify the console operator that you are prepared to lower the CET into the core; when acknowledged, lower the CET into the core ensuring that it is properly seated in the lower grid plate.

k. With a downward pressure on the CET to keep it seated, secure the CET bracket with the two bolts,

l. Ensure appropriate entries are made in the operations logbook and the fuel book, and that the reactor core pegboard is updated.

2. Irradiation:

a. Clean the rabbit (s) using alcohol and water.

I b. Once clean, do NOT handle the rabbit e:: ept with gloves, Kimwipes, or handling tools.

c. Ensure that the rabbit cap is secured tighte.

d. Bring the reactor up to the appropriate power.

e. After notifying the reactor operator on console, drop or lower the rabbit into the core WITH THE cap UP. Ensure that this individual spends a minimum amount of time in the vicinity of the carriage. Do NOT lower the rabbit with the extractor tool while at power.

f. Complete irradiation and shut down reactor.

g. Ensure appropriate entries are made in the operationr, logbook and the CET logbook.

3. Rabbit Retrievals:

a. Ensure that a reactor staff member and a Safety & Health Department (SHD) monitor are present in the reactor room and that they are wearing all required (whole body TLD, pocket chamber and wrist) dosimetry.

If the CET is in the core, a reactor operator must monitor the console during the retrieval.

b. Test the rabbit extractor (" fishing pole") for operability.

c. Insert the extractor head mechanism into the CET and reel out ceble until you reach the low end indicator painted on the cable.

I

%h_ _ _ . _ _ _ _ . . _ _ . _ _ _ _ _ _

1 l

i

d. Drop the extractor head firmly on the rabbit.

e. Ensure the SHD monitor has a teletector positioned near the CET 4 top to monitor the rabbit. l

f. If the CET is in the core, notify the reactor operator that the j' rabbit is being pulled and continue when acknowledged.

g. Reel in the cable at a rato commensurate with radiation levels; lower the rabbit back into the CET if the rabbit is excessively hot,

h. Stop when upper end indicator is visible on the cable; have SHD q take an accurate radiation reading. l l

i. If radiation levels are acceptable, swing rabbit away from carriage i I

and have another individual grab it with a handling tool. If the radiation I levels are not acceptable, lower the rabbit back into the CET. The rabbit will l again be withdrawn for reevaluation of radiation levels when the SHD and RFD representatives concur on an acceptable radiation level in accordance with ALARA and mission requiremt nts.

j. Release extractor head and detach rabbit from head,

k. Unless working with the rabbit, or radiation levels are very low K1 mR/hr), store rabbit or irradiated material in a lead pig or storage cask,

i. Make appropriate entries in the operations and CET logbnoks.

4. CET Removal from Core:

a. ' Complete steps la-c above.

b. Loosen the CET bracket. bolts while holding the CET down; remove the CET bracket,

c. Notify the console operator that you are prepared to remove the

. CET from the reactor core.

d. When acknowledged, transfer the CET to the storage rack, ensuring that it is kept as low in the water as possible.

e. Secure the CET with cable ties,

f. Secure the CET bracket with the two bolts.

g. Remove the fuel element from the storage rack and transfer to core. Notify the console operator and receive acknowledgment prior to insertion of element into fuel position F28.

h. Ensure the element is properly seated in the lower grid plate by listening for the " double clicks".

i. Make appropriate entries in the operations and fuel logbooks and update the reactor core pegboard.

m _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . . _ _ . - _ . _ _ _ - _

'\

Revised: Dec'1987 TAB C . EXTRACTOR SYSTEM GENERAL: The extracte system will be tested for operability prior.'to the

' initial' experiment for the day..

SPECIFIC:

- 1. Assembly of the extractor system:

a. Inside the exposure room:

(1) Move the inside receiver section into position in front of the core; screw, tube supports to the floor and place lead bricks on them.

(2) ' While holding the appropriate connecting tube in position, tie the

~'

strings in the tube to. the .two ends coming out of the exposure room wall and

. to the two ends in the receiver section.

(3) Align the ends of the tubes and slide the clamp over each joint.

(4)' Place the alignment tools into the appropriate holes to check the tube alignment; tighten down the clamps.

(5) Connect the electrical cable to the limit switch.

(6) . Remove the' alignment tools.

b .~ Outside the exposure room:

(1) . Remove tube plug.

(2)- Move the receiver section close to the tube projecting from the wall. :

-(3) Tie the string from the end of the small tube to the end of the

' wire cable.

(4) Pull the. string in the large tube slowly while having someone inside the room guide the string.

(5) When the cable -is all the way- through both tubes, thread the cable 'through .the receiver tube while moving the receiver table into final position against the ' wall (if necessary, add an additional length of cable to the take-up reel).

(6) ' While someone else is pushing the table toward the wall, insert two screws into the holes on the securing bracket (beneath the table).

(7) . position and tighten clamp over the joint;. position carrier in tube and connect cable to each end; removethe tape on the take-up reel.

(8) Pull- back on the drive motor assembly until there is no slack in the cables; tighten the adjustment bolts on the drive assembly.

(9) Connect the electrical cables to the ' motor, control unit, and limit switches.

2. Disassembly:

a. Reverse the order of the above with the following changes:

(1) Before loosening the motor assembly, place tape on the cable drum to keep the cable from moving (ensure the carrier is in the receiver section).

-(2) Before pulling the cable.through the tubes, attach a new string to it.

(3) . Leave enough slack for disassembly inside the exposure room.

(4) Cut the string at the joints in the room and tape the ends to the tu be s.

b. Ensure the tube plug is in place, and the control unit is secured.

______.E._____.-___________-

i L l l

h

! 3. Operations:

a. On the motor control, initially set controls as follows:

l (1)' Power switch: "OFF". 1 (2)- Torque ' control: "OFF". ~1 (3) In/out. switch: " BRAKE".  !

,. (4) Speed control: "0%". ..

{ b. plug motor control into AC outlet, switch the power switch to "ON".

c. Switch in/out switch tc appropria*,e position.

d. Slowly increase- speed to an appropriate level; as the carriage approaches its full in/out position,' decrease the speed slowly to ."0%"..

j e. Turn the in/out switch to " BRAKE". q

f. During power operations, ensure that the following requirements are '

, met:

! (1) The prep area is sealed off.

(2) A Safety & Health Department (SHD) moniter is present.

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._______m.._ _ _ _ _ _ _ _ _ _ _ _ _ _

m.

k L Revised: Jan 1984 o

! TAB D . PNEUMATIC TRANSFER SYSTEM (PTS)

L.

General:

1. This (PTS) procedure is inactive. If the PTS Facility is reactivated, then this procedure must be reviewed and approved by the RRFSC and the Reactor r Facility Director.

l

2. ALARA principles will be practiced during PTS operations.

3. All PT1 operations will be directly supervised by a reactor operator present in the Hot Lab.

! Specific:

1. PTS Setup:

a. Position core at 833 (Inside region III).

b. Ensure communications are established between the hot lab and the control room.

c. . Inspect rabbits to be used in the PTS for cracks or other damage.

d. Aluminum rabbits must be diverted to the Hot Cell and therefore may only be used on the "A" system.

e. If the anticipated radiation level of any returned rabbit is greater than 1.0 R/hr at 1 meter, take the following precautions:

(1) Use the remote control unit, unless experiment requirements

' dictate .otherwise.

(2) Place a radiation survey meter next to the receiver / sender station so that it can be monitored from the remote control unit.

(3) The rabbit will be irradiated in the "A" system and then diverted to the Hot Cell or returned to the irradiation location.

2. Manual Operations:

a. Ensure all switches on both the local and remote control units are in the "OFF" position; place the local / remote switch in the desired position.

b. Place blower switch in the "ON" position.

c. Insert key ~ into local control unit; turn key to "ON" position.

-d. Ensure tubes are empty.

e. Set mode switch (man /off/ auto) to " MAN" position. Blower will start.

f. Set in/out switch to the "OUT" position and the tube on/off switches to "ON"; allow the -system to run for a short time.

g. Set tube on/off switches to "OFF" and turn in/out switch to "IN".

h. Load samples into tubes.

i. Check communications with reactor operator at the reactor console.

J. When the reactor. is at the designated power level, set the tube on/off switches to "ON" one at a time, to send rabbits into the irradiation location,

k. Begin stopwatch or timer.

1. Turn tube on/off switches to "OFF" and turn in/out switch to "OUT".

m. Ensure a Safety & Health Department (SHD) monitor is present during retrievals.

n. Set on/off switch to "ON" one at a time; rabbits will return to sender / receiver station.

o. Set all switches to "OFF", and remove key from control unit.

,\

l

! 3. Automatic Mode:

a. Complete steps 2a-d above.

b. Set mode switch to " AUTO" position. Blower will start.

l c. Complete steps 2f-i above.

i d. Set timer (0 to 5 minutes) by turning the red and black arrows to the desired irradiation time.

) e. When the reactor is at the desired power level, briefly push the timer l push botton and release. The rabbits will leave the receiver / sender station and will automatically return at the end of the preset irradiation period. The timer will automatically reset.

l f. Turn all switches to "OFF" and remove key from control unit.

4. Diverting Samples:

l a. Diversion of samples to the Hot Cell may only be made using the 'A" system,

b. Af ter the rabbit has returned to the receiver / sender station, set the divert / send switch to " DIVERT" and hold it until the loading port handle trips to the rear position,

c. Send the divert / send switch to " SEND" and hold for a few seconds.

The rabbit will ' leave the receiver / sender station and travel to the Hot Cell.

1 l

1 I

Revised: 15 Dec 88 TAB E IN-POOL /IN-CORE EXPERIMENTS General:

ALARA principles will be followed during these experiments. These procedures apply to all in-pool or in-core experiments except CET operations (See Procedure 1 - Tab B).

Specific:

1. All operations will be supervised by an SRO.

2. Actions will be taken to prevent damage to the reactor core or aluminum tank.

3. Ensure that a member of the reactor staff and a SHD representative are present during the removal of samples from in-pool or in-core locations.

4. The removal of experiment materials from the pool or core will be monitored with a radiation survey meter; additionally, a reactor operator will monitor the reactor console during insertion and removal of in-core experiments.

I I

I I

I I

I I

E .

i

l- '

! l Revised: Jan 1985 PROCEDURE II REACTOR STAFF TRAINING

1. The reactor staff training is delineated in the current "AFRRI Reactor Operator Requalification Program". I i

2. The Reactor Facility Director (RFD) determines who is allowed into the training program. As part of the training /requalification program, the following will be performed:

I a. A training file will be maintained for each trainee / operator.

b. When a section of training is completed, it will be annotated on the training checklist in each file.

c. A record of operations will be kept for each trainee / operator.

I I

I I

I I

I

Revised: Jan 1985 PROCEDURE III MAINTENANCE PROCEDURES General: Maintenance procedures are provided in other references.

S pecific:

1. Preventative Maintenance procedures for each itera of the reactor systems are provided in the maintenance logbook.

2. Annual shutdown procedures are given in the Annual Shutdown Checklist which is revised each year by the Reactor Operations Supervisor (ROS) and approved by the Reactor Facility Director.

3. Malfunctions are annotated in the Malfunction Logbook. Each entry is made by the operator who discovered the deficiency. When corrective actions have been made and annotated in the malfunction logbook, the RFD or ROS shall review and initial the entry.

4. Procedures for maintenance of specific equipment are provided in the manufacturers' literature.

I I l l

l I

I I I

g; l

1 I

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

l f.

Revised: Jul 1982 PROCEDURE IV PERSONNEL RADIATION PROTECTION 1

1

) General: All activities performed in areas of potential personnel radiation )i exposure will be done in accordance with ALARA principles. These areas are I the reactor room, upper equipment room (3152), lower equipment room (2158), I

! warm storage, prep area, exposure room 1, exposure room 2, and the hot i lab / cell.

Specific l

l

1. Reactor Room: ]

a. CET Operations: See Procedure I-Tab B. 1,

b. Working inside chained in area around pool: The reactor operator on the console shall be responsible for controlling entry into the chained area around the pool.  !

2. Warm Storage: See HPP 3-3. l l

3. Prep Area: See Prep Area Briefing.

4. Exposure Rooms: See HPP 3-1 and Procedure I-Tub A.

5. Hot Lab / Cell: See HPP 3-5 and Procedure I-Tab D.

6. Upper and Lower Equipment Rooms:

a. No written radiation protection procedures are required for entry into these rooms.

b. However, access to these areas is controlled by the AFRRI Reactor Physical Security Plan.

7. Personnel Dosimetry and Monitoring: See HPP 3-1, 3-2, and the Prep Area Briefing.

i l

I I

1 i

J

t Revised: 15 Dec 88 j l

PROCEDURE V PHYSICAL SECURITY General:

Physical Security requirements are given in the AFRRI Reactor Physical Security Plan.

I Specific:

1 j

1. The reactor control room and the reactor room will be secured if no reactor staff member is present for a prolonged period of time during duty hours.

2. Control of keys is delegated to the Reactor Operations Supervisor, Key inventories will be performed annually, not to exceed 15 months.

i I  !

I I

I .

B 6

d Revised: 15 Dec 1988 PROCEDURE VI EMERGENCY PROCEDURES General: The reactor emergency organization, emergency classes, and emergency action levels are set forth in the current copy of the AFRRI Reactor Emergency Plan.

S pecific: Perform the following, as appropriate (need not be done in orders.

1. Reactor Emergency:

a. SCRAM reactor, I b. Check radiation monitors; use portable survey instruments to assess situation, if necessary.

c. Notify ERT Commander of situation.

d. Activate emergency organization.

2. AFRRI Complex Emergency Evacuaticn:

I a. SCRAM reactor.

b. Secure any exposure facilities which are in use so that personnel access to that facility is not possible.

c. Remove logbook, emergency guide, radios, teletector, tool kit, and keys; I repor* to EAS.

d. Do NOT lock reactor area doors.

I l I 1

)

1 I  :

l 1

I l I .

f Dec '87 PROCEDURE VII REACTOR CORE LOADING AND UNLOADING l

General: Loading and unloading of the reactor core shall be under the supervision of the Reactor Facility Director or the Reactor Operations Supervisor. These procedures are superseded in the following situations: during CET Oper-ations (see procedure I-Tab B) and during annual shut-down maintenance (see the current Annual Shutdown Checklist).

Specific:

1. Setup

a. Ensure at least one nuclear instrumentation channel is operational. i

b. Ensure an operator monitors the reactor console during all fuel movements.

c. Check new fuel elements prior to insertion into the i core; this includes cleaning, visual inspection, and  !

length and bow measurements.

d. If irradiated fuel elements are to be removed un-shielded from the pool, a Special Work Permit (SWP) will be obtained from the Safety & Health Department (SHD); fuel elements with a power history (greater than 1 KW) in the previous two weeks shall not be removed from the reactor pool.

2. Core Loading

a. After each step of fuel movement perform the following' l

(1) Record detector readings.

(2) Withdraw control rods 50%; record readings.

(3) Withdraw control rods 100%; record readings.

(4) Calculate 1/M.

(5) Flot 1/M versus number of elements (and total mass of U-235).

(6) Predict critical loading. ]

(7) Insert ALL rods; continue to next step. {

b. Load elements in the following order: J l

(1) Load the "B" ring thermocouple element. {

(2) Load the "C" ring thermocouple element. )

(3) Install temperature measurement system (to measure fuel temperature.

L (4)

Install any other thermocouple elements.

(5) Complete loading of "B" and "C" ring elements (total of 18 elements).

(6) Load "D" ring (total of 33 elements)

(7) Load the following "E" ring elements:

1,2,4,6,8,9,10,12,14,16,17,18,20,22,24 (total of 48 elements).

(8) Complete the "E" ring (total of 57 elements),

(9) Load the following "F" ring elements:

1,5,9,13,17,21,22,23,27 (total of 66 elements).

(10) Load two elements per step until critical loading is achieved.

(11) Load core to $2.00 excess reactivity.

(12) Estimate control rod worth using rod drop techniques.

(13) Estimate the control rod worth of the remaining unloaded elements.

(14) Load the core to achieve a K-excess that will allow calibration of the TRANS rod based on the  !

last available worth curve of the TRANS rod. l (15) Calibrate the TRANS rod.

(16) Estimate the shutdown margin. I (17) Estimate K-excess with a fully loaded core (must l not exceed $5.00).

(18) Load core to fully operational load and recalibrates all control rods.

3. Core Unloading:

a. The reactor core will be unloaded starting with "F" ring and ending with the "B" ring.

b. The fuel elements will be individually removed from the reactor core, identified by serial number, and placed in either the fuel storage racks or a shipping cask.

c. If elements are to be loaded into a shipping cask, perform a complete cleaning of the cask and check for radiological contamination prior to placing the cask in or near the pool. Load cask in accordance with procedures specific to the cask.

d. Once the cask is loaded, perform an air sample and survey; check temperature and pressure inside cask, if necessary.

e. If elements are placed in temporary storage away from core monitoring, insure criticality monitoring in accordance with 10 CFR 70 is in place.

Revised: 15 Dec 1988 PROCEDURE VIII REACTOR OPERATIONS General:

Logbook entries will be made in accordance with the Logbook Entry Checklist (Tab A).

Specific:

1. Each line on the daily and weekly checklists shall be initialed by the Reactor Operator or Trainee who performs that item.

I. 2. Perform reactor Daily Operational Startup Checklist (Tab B), utilizing appropriate nuclear instrumentation set points (Tab C). In the case of no planned operations, a Daily Safety Checklist may be performed (Tab B1).

I 3. Record at the top of each page the SRO On-Call for that date.

4. Perform K-excess measurement (Tab D).

Perform operations in accordance with the following:

5.

a. Steady state operation (Tab E).

i

b. Square wave operation (Tab F). l

c. Pulse operation (Tab G).

d. CET operations (Procedure I, Tab B).

I 6.

e. Pneumatic Transfer System (Procedure I. Tab D).

Perform Weekly Operational Instrument Checklist once during calendar week (TAB H).

7. At the end of each day in which a Daily Operational Startup Checklist or Daily Safety Checklist has been completed, perform Daily Operational Shutdown Checklist (Tab I).

8. Complete the monthly summary (Tab J).

I I

I I

I .

Revised: 15 Dec 1988 l

TAB A LOGBOOK ENTRY CHECKLIST

, 1. The reactor operations logbook is a before-the-fact record, that is, entries l' will be logged before the operator actually performs the planned function.

Any late entries will be so no' l

2. The operations logbook will have a hardbound cover and will be sequentially numbered by volume. The pages will be dated at the top of each .

page and each page will be sequentially numbere 4. ]

3. The Reactor Facility Director (RFD) will review each logbook upon its completion; he will make an appropriate entry in the back of the logbook and  :

sign the entry.

4. The entries will be made in ink and in accordance with the following )q

' designated color code:

l

a. Black and Blue-Black:

(1) Console locked and unlocked. The individual at the console will enter his/her name and the supervisory licensed operator's name, if j necessary. l (2) . Checklist number and completion time. l (3)' power level at criticality and subsequent power level changes. J (4) Reactor SCRAM. l (5) Mode of operations. Use appropriate stamp or entry to designate '

the operation:

(a) Mode I or IA Steady State (b) . Mode II Square Wave (c) Mode III Pulse (6) Operation of reactor associated facilities such as lead shield doors, pneumatic tube systems, etc., unless such operations cause a change of ,

I reactivity (see 4.b.(2) below).

(7) Change of personnel at the console. Name of personnel will be entered along with the licensed operator present in the control room, if the person at the console is not a licensed operator.

(8) The operator in charge will be designated in the logbook  ;

whenever multiple operators are signed on the console. ]

(9) Completion of the daily startup and shutdown checklists, and J week!r checklists. {

(10) Signature of reactor operator to close out the log for the day.

'. (11) Reactor calibrations and data.

I

b. Red. {

(1) K-excess measurements, to include experiment worth determinations.

(2) Actions which affect reactivity:  ;

(a) Core movement. )

(b) Fuel movement. l (c) Control rod physical removal for maintenance. I (d) Experiment loading and removal from the CET, PTS, pool, or I core.

l l

c. Green. j (1) Reactor malfunction,to include the reactor systems and eupport j equipment taken out of service for maintenance and returned to service. I (2) Additional items entered at the discretion of the operator such as I addition of makeup water to the reactor pool, etc.

5. When an operation requiring entry into the logbook falls under more than one color code, the color to be used will be determined via the following order of precedence: RED - GREEN - BLACK / BLUE-BLACK.

1 1

I i

t I 1 l

l I

i l

I l l

l 1

l l

I l l

.5 TAB B: DAILY OPERATIONAL START UI' CHECKLIST Checklist

• Date SENIOR SRO PRESENT/ON CALL Performed by Operators Time Completed I. EQUIPMENT ROOM (ROOM 3152)  !

I 1. Air compressor pressure (psi) .....................

2. Air compressor water trap drained .................

3. Air dryer operating ...............................

4. Doors 231, 231A, 3152 and roof hatch secured ......

II. LOBBY AREA Lobby Audio Alarm turned OFF ......................

III. EQUIPMENT ROOM (ROOM 2158)

I 1. Prefilter differential pressure ...................

2. Primary discharge pressure (psi) ..................

3. Demineralized flow rates set to 6 GPM . . . . . . . . . . . . . _ .

.4. Stack roughing filter 6p (inches of water) ........

5. Stack absolute filter 6 p (inches of water) ........

6. Visual inspection of area .........................  ;

7. Door 2158 SECURED .................................

IV. PREPARATION AREA Visual inspection of area .........................

V. REACTOR ROOM (ROOM 3161)

. . - - - _ . _ . . - - . - _ _ . _ .-. l I 1.

2.

3.

Transient rod air pressure (psi) ..................

Shielding doors bearing air pressure (psi) ........

Tank water level below full mark tinches) .........

l

4. Visual inspection of core and tank ................

I 5. Number of fuel elements and control ... fuel elements rods in tank storage control rods j

l

6. Air particulate monitor (CAM)

(a) Operating and tracing .........................

(b) Alarm test complete ...........................

7. Door 3162 SECURED ................................. }

8. Stack gas monitor quality assurance checked .......

1 AFRRI FORM 61a (R) Revised 26 Apr 88 I Reformatted 18 Aug 88  ;

l 6

l

VI. REACTOR CONTROL ROOM I 1. kmergency air system RESET .......................

2. Console recorder dated ............................

3. Stack gas and fuel temp. recorder dated ...........

4. Logbook dated and reviewed ........................

I 5. Water monitor box (all conductivities must be > 0.5 M -em)

(a) Background activity (mA).......................

(b) Alarm test completed and alarm reset to 0.5 mA (c) Water monitor box conductivity [MG-cm) ........

(d) DM1 conductivity [MG-cm] ......................

(e) DM2 conductivity (bu-em] ......................

6. Stack gas flow rate [Kefm] ........................

7. Stack gas monitor (a) Background (cpm) ..............................

c) H gh ala m set to 8bb "$1PC kr-41 . .....l.....

8. Stack particulate monitor (a) Background (cpm) ..............................

(b) Alarm check ...................................

(c) High alarm set to 2.0 E3 cpm ..................

9. Radiation monitors Alarm Point Reading Alarm Setting I Monitor (a) R-1 Functional (mR/hr) (mR/hr) 500 i

(b) R-2 10 I (c)

(d)

(e)

R-3 R-5 E-3 10 50 10 (f) E-6 ,

10

10. Timer on .........................................

11. TV monitors and beeper system on .................

12. Source level on log channel > 0.5 Cps ............ j

13. Check 0.5 Cps source RWP .........................

14. Time delay operative .............................

15. Operational channels i I

I Cal, log switch Pos. # Range Switch  % Linear  % Log 1 0.3 natts 2 30 watts I 3 300 watts 4 1 kw 5 10 kw 6 3 Mw

.16. Rod raising interlock for mode I .................  !

17. Rod raising interlock for mode III ............... l

18. Zero power pulse (obtain signal w/ trip test) .....

19. SCRAM checks (at least one per rod)

(a) Safety flux 1 ... (g) Emergency Stop .. 1 (b) Fuel temp 1 ..... (h) . Pool H:0 level .. l' (c) HV loss safety 1 (i) Fuel temp 2 .....

(d) Manual .......... (j) Safety flux 2 ...

(e) Reactor key ..... (k) HV loss safety 2 (f) Timer ...........

I 20.

21.

22.

Water temperature (inlet) ........................

Period trip test for 1 kw interlock ..............

CAM high level audible alarm check ...............

b

• - TAB B1: DAILY SAFETY CHECKLIST v F

Checklist e DATE SEN!OR SMS PRESENT/ON CALL PERFORMED BY OPERATORS TIME COMPLETED I 1. EQUIPMENT ROOM (RM - 3152)

I 1. Air compressor pressure (psi) .........................

2. Air compressor water trap QRA,,LN1Q .....................

3. Air Dryer Ocorating ....................................

j Doors 231.231A. 3152. and rooi hatch SECURED ..........

4 i

1 l

II. LOBBY AREA Lobby Audio Alarm turned E ..........................

III. EQUIPMENT ROOM (RM-2iSS)

1. Profilter differential pressure .......................

2. Primary Discharge Pressure (PSI) ......................

3. Domineraliser flow rates set to e gpa .................

4. Stack roughing filter 4p (inches of water) ............

S. Stack absolute filter 4p (inches of water) ............

I 6. Visual inspection of area .............................

7. Door 213G M .....................................

IV. PREPARAT!QH AREA Visuai I n s pec t i on o f A r e a . . . . . . . . . . . . . . . . . . . . . . . . . . . . . {

I I

1 I

O

I Apr '88 l

' TAB C NUCLEAR INSTRUMENTATION SET POINTS General: These set points may be adjusted for a specific operation by of the RFD or ROS but in no case may they be set at a point non-conservative to the technical specifications.

Specific: The following are channel or monitor set points (alarm, scram, rod withdrawal prevent).

1. Scrams:

I-a.

b.

c.

Fuel Temperature 1 & 2:

High Flux 1 & 2:

Safe Chambers 1 & 2 HV Loss:

575 C 110% (1.1 MW)

Loss of 20%

d. Pulse Timer: 0.555 seconds

e. Steady State Timer; as necessary

2. Rod Withdrawal Prevents:

I a. Period: 3 seconds

b. 1 KW (Pulse Mode): 1 KW

c. Source: 0.5 CPS

d. Water Bulk Temperature: 50 C

e. Fission Chamber HV Loss: 20%

3. Alarms:

a. RAMS: As directed in procedures !

b. CAMS: 10,000 CPM c .- Stack Gas: 800 MPC Ar-41

d. Stack Particulate: 2.0E+3 CPM I e.

f.

Water Monitor Box Gamma:

Criticality Monitor (RS):

0.5 mA 50 mR/hr day -

20 mR/hr night or as directed I

I 4

l l

I Jul '82 l

l TAB D K-EXCESS j l

l

1. Withdraw SAF and SHIM rods 100% and withdraw the TRANS rod {

25%. l 1

1

2. Une the REG rod to bring the reactor to cold critical at 15 j watts. If criticality can not be reached with the REG rod I full out, use the TRANS rod to bring to critical.

3. When power is stabilized at 15 watts, record rod positions in -

reactor operations logbook, entering all information in red ink.

3 4. Using rod worth curves, compute K-excess for the core g position

• used and record in the reactor operations logbook and on the Monthly Summary Sheet. <

• Note: Use the curves for position 567 when doing K-excess at I 290. J 1

I I l 1

I I

I C

" Jan '85 l TAB E: STEADY STATE OPERATION (MODE I/IA) {

lI

, General: The reactor shall not be operated at a power greater i

i than 1.0 MW. i Specific:

1. Set the linear channel range select switch to the appropriate scale for the desired power (at this power the linear pen should be as close to 50% as possible).

2. Set the mode switch to manual mode.

3. Raise control rods with the appropriate banking, taking into consideration the location in the pool, power level, and experiment array.

4. If final approach to critical is to be made in automatic mode, I perform the following:

l

a. Set the proper percentage on the flux control to obtain the desired power.

b. Raise the TRANS, SAFE, and SHIM rods to the appropriate banking.

c. Raise the REG rod approximately 5%.

d. When the servo has locked in, switch to automatic mode.

I

e. When the reactor reaches critical, fine tune the flux control for the desired power.

5. Scram the reactor at the end of the run using the manual or j timer scram.

6. Ensure the appropriate entries have been made in the opera- i tions Jogbook. {

I' Note: For runs greater than 800 KW, adjust alarm points on R-1

{

and R-5 to full scale.

II Apr '88 TAB F: SQUARE WAVE OPERATION (MODE II)

General: The square wave mode can not be used above 900KW.

Specific:

1. If appropriate, set timer for run duration and flip timer SCRAM switch to "ON".

2. Set the TRANS drive anvil as follows: >

a. If the TRANS rod is not to be used for criticality, use the applicable rod worth curve to determine the anvil position for a 75 cent insertion and raise the anvil to that position. Adjust the REG rod to 90 percent; achieve criticality using the SHIM and SAFE rods.

b. If the TRANS rod is to be used for criticality, withdraw I the REG rod 90 percent, SAFE and SHIM rods 100 percent, and adjust to critical configuration using the :TRANS rod.

When a critical configuration has been reached, drop air from the TRANS rod and adjust the anvil position for a 75 cent insertion above the critical position.

3. Adjust power range select switch to the desired range.

4. Set square wave percent demand dial to 80 percent of final desired power.

5. Set flux controller dial to final desired power level.

6. Switch into square wave mode, making sure the TRANS rod ready light is "ON".

7. Depress ready / fire button.

8. After the servo has locked in, raise the square wave percent demand dial setting to the final desired power.

9. Switch to manual mode and lower REG rod to 80 percent while raising the TRANS rod; maintain desired power level and then switch to automatic mode.

10. Scram the reactor with the timer or manually, as appropriate; move the core if applicable.

11. Ensure all pertinent information has been logged in the reactor operations logbook.

May 88 (Pending RRFSC Review)

I TAB G: PULSE OPERATION (MODE III)

General: Pulses above $3.50 must be approved by the RFD or ROS.

Specification en the RUR may be used to meet this requirement.

For pulses fired from COLD CRITICAL, omit steps 2, I. Specific:

6, and 8. For pulses fired from SUBCRITICAL, omit 5,

step 7.

1. Set the alarm points on R-1 and R-5 (the criticality monitor) to full scale.

2. Given a core position, set the transient rod at a position corresponding to the dollar value determined by the following equation:

$ Value = Total worth ($) TRANS rod - desired pulse ($) value

3. . Bring the reactor to cold critical (15 watts) with the three I standard rods, using a rod configuration commensurate with core position.

4. Stabilize in manual mode.

5. SCRAM the transient rod.

6. Raise the transient rod anvil to 100%.

7. Raise the transient rod anvil to the desired pulse position.

8. Let the power decay to approximately one watt.

9. Place power range select switch on the "3 MW-PULSE" position.

10. Place mode select switch in "?ULSE HI" (greater than or equal to $2.15) or " PULSE LO" (less than $2.15) position, as appro-I priate. i

11. Fire pulse by depressing " READY" button on the console (Note:

Power must be below 1 KW).

I

i

12. Record data in the reactor operations logbook as follows:

Insure the pulse detector selector switch is on

'I a.

Detector #2.

b. Turn reactor pool lights and reactor room overhead lights OUT.

c. NVT: Read from Safety Channel 2/NVT meter on right side of console. Multiply the Safety Channel 2/NVT meter reading by 1.45 to obtain actual NVT reading in MW-s.

Full scale is 43.5 MW-s.

d. NV: Read blue pen on chart recorder, center of console with the following equations:

1. Pulse Hi switch: 2500 MW X (% of scale reading) X 1.45 I

2. Pulse Lo switch: 500 MW X (% of scale reading) X 1.45

e. Fuel Temperature: Read red pen of chart recorder, center of console, where 100% = 600 C.

13. Reset mode select switch to manual mode after reading are recorded ,

l

14. Reset R-1 and R-5 to their normal alarm points.

4 I I I

I

r WEEKLY O PER A TIO N AL IN S T R U V1 E N T CHEC hs7 k

~

CHECKLIST No. DATE g T PERFORMED 8Y _ REVIEWED BY

1. SAFETY CHANNEL ONE A, Raise rod m2%, depress and reisese switch marked MV # 2 in left nand drawer, OD$srve and reset scram .

8. Rotate coerate switch to zero, check meter for zero, reset scram __

C. Rotate coerate switch to calibrate, check meter for 100 %, reset scram

11. OPERATIONAL CHANNEL HIGH VOLTAGE Decress and hold in switch HVJt 1 in left nand drawer Attemot to raise control rod 111. SAFETY CHANNEL TWO A. Aaise rod ~2%. scoress and release switen -V tr'o test in rignt mand : rawer.

Cesene and reset scram .

8. Actate ccerate sWiten to zero. chec= '"eter Por tero, reset sc'a"1 C. Aotate ecerate switen to calierate. cnec= aeter 'er 100 %. reset scram IV. NV.NVT A. NV. In manual moce, set 20% on safety channel as 2 with tric test knoo. Switch to pulse Hl, chart recorder should read 20%. Repeat for 40%. 60%. 80%. and 100%. Check scram at 110% .

8. NVT Check NVT circuit by procedure 3.3.5.2 through 3.3.5 A in console maintenance manual.

V. FUEL TEMPERATURE NO.1 Rotate switch to estibrate position observe 100% on meter reset scram . .... .. ..

VI. FUEL TEMPERATURE NO. 2 Rotate switch to calibrate position, observe 100% on meter reset scram .

Vll. WATER LEVEL INDICATOR A. In cool, east side, depress float on water level indicator . .

B. Observe scram on console. (scram indication should reset automatically) .

Vill. WATER CONDUCTIVITY List resistivity readings for previous week from daily startuo checklists. Determine that average at each point is greater tran MON TUE WED THU FRI A VE Monitor Box _

DM1 OM2 IX. RADI ATION MONITORS A. Test alarm functions for high level and failure HIGH Level Monitor failure Alstm functional Alarm functional R1 R .2 E.3 E6 R 5 (criticality Reactor Rm APM Gas Stack Monitor B. Reset Alarms APRRi PORM te IPREVIOUS EDITl0848 OP THIS 70Rea ARE OSSOLETE.)

19 OCT Sa

7 d

i l

l

. DAILY OPERATIONAL SHUTDOWN CHECKLIST 1

I Checklist No.

Time Completed Date Performed by 1 l

I. REACTOR ROOM (Room 3161)

1. All rod drives DOWN ...............................

2. Carriage lights OFF ...............................

3. Door 3162 SECURED ........................... .... j

4. Door 3161 locked with key .................... ....

5.

I Print out an hourly report from the Stack Gas Monitor ......................

II. EQUIPMENT ROOM (room 3152)

1. Distillation unit discharge valve CLOSED ..........

2. Ai. dryer OPERATIONAL .............................

3. Doors 231, 231A, 3152 and Roof hatch SECURED I ......

E III. EQUIPMENT ROOM (Room 2158)

1. Primary discharge pressure (PSI) ..................

2. Deminerelizer flow rates set to 6 GPM .............

3. Visual inspection for leaks .......................__ )

4. Dorr 2158 SECURED .................................

I IV. PREPARATION AREA 5 1. ER 2 plug door CONTROL LOCKED; E Door closed; and handwheel PADLOCKED ............

2. ER 2 lights ON and rheostat at 10% ................

3. ER 1 plug door CONTROL LOCKED:

Door closed; and handwheel PADLOCKED ............

4. ER 1 lights ON and rheostat at 10% ................

5. Visual inspection of area .........................

AFRRI FORM 62 (R) Revised: 26 April 1988 I Reformatted: 5 December 1988

%-~m_m.________ _ _ _ _ . _ _ _ . _ _ . . _ -

V. LOBBY ALARM

1. Lobby alarm audio ON ..............................

I VI. REACTOR CONTROL ROOM (Room 3160)

1. Reactor tank lights OFF ..........................

2. Timer OFF ........................................

3. TV monitors OFF ..................................

4. Console LOCKED ...................................

5. Diffuser and secondary pumps OFF .................

6. purification and primary pumps ON. ...............

~. Beeper system turned OFF .........................

8. Reactor monthly usage summary completed ..........

9. Exposure room camera power supply turned OFF .....

10. Radiation monitors . .............................

MONITOR READING HIGH LEVEL ALARM SETTING (Mr/Hr)

a. R-1 20

b. R-2 N/A

c. R-3 N/A 20 I d.

e.

f.

R-5 E-3 E-6 N/A N/A

g. R-6 N/A I 1 I

4 I  ;

I 1

l l

l l

1 l

l l

l

i McAL I UM MUN f HLY USA ESUM'/A g.gxgE33 muugE4 4Ut!8ER PVLSES , *0$17104 $7EADY STa rt.P',;dE3 f0TAL TOTAL , a;At v

Af START. Op 2 as gtgg Upt E A NO.1 CENTE A TUIES E A 40J OTHE A STEADY 'U L3E I *cwtA PUL188 E2 3 (231) (147) (8338 (M31 ( ) Smuf-STATE PowlR 833 TOTAL TOTA 4, TOTAL TO 00*ms w TOTAL 8 imusamsman OATE TOTAL name====

I CARRY 4VE R DATE l

l l Daft I I

I l l ' l I l l l OAft I I ' I '

I '

l I i i +

1 I DATE CATE I I '

I t

i I

I i

I t i i ,

OAM l I i l I ( l i I i .

l

~

l

~ 9 ~

0 ATE  : I i t I I GAM I

i k_ ___

t I

.0 ATE l l ' t l l

! l i '

I , i i

OAM l l l l i l 1

I OAM [ l 8

i I

l I

i i  !

l I

; i DAM l l l ' '

I .

OAM l

l l

I i

? .

i t

l . .

CAM l I '

I OATE I

l l

i f

I OATE OAft I

l l

l

_ {

l l

l i

1 I .

OATE I

I l

I

\

l l

I I DATE DATE l l l

l

_ _I

[NYd" " .

MONTM

I t

\

Dec '87 I TAB K: STACK GAS MONITOR PROCEDURE 1 l

GENERAL: This procedure specifies all the requirements for operation of the Stack Gas Monitor (SGM) in the reactor room.

This instrumentation is used to monitor, measure, and record j the amount of Argon-41 released to the environment through a the reactor stack. I l

SPECIFIC: l

1. QUALITY ASSURANCE PROCEDURES: A quality assurance check I is performed daily, prior to reactor operations, as part of the reactor start-up. This check is performed in the following manner:

a. The particulate filter is changed and the old fil-ter is discarded as radioactive waste.

b. The date and time of this quality assurance check is entered in the log.

The Detector voltage system set point is checked and I

c.

recorded in the log: q

d. The air sampling flow rate (should be greater than 3.5 cubic feet per minute) is recorded in the log.

e. The front cover of the detector shield is removed j and the check source is inserted all the way in to i the face of the detector. The blue alert light I shruld come on as the count rate rises above the alert setpoint. The red high level alert light I and bell should come on as the count rate rises above the high level setpoint. The audible alarm can be silenced by pushing the red button on the front of the SGM cabinet.

f. After 5 minutes have elapsed, an historical one I minute report printout should be made and the counts per minute for the third minute of high level readings should be recorded in the log.

l

g. The alarms should be acknowledged in the log with the initials of the person performing the check and on the SGM by pushing the " acknowledge" button on the keyboard.

h. The check source should be removed and the detector I cover should be replaced. l l

l l

ll .

1

l

, )

6 I 1. Finally, the counts per minute reading (paragraph 2f) should be checked against the plot of counts per minute versus Julian date to determine if it falls within a plus or minus 5% deviation for the detector and check source. This check provides the j necessary quality assurance fot> the SGM system prior I to conducting any reactor operations for the day.

2. HISTORICAL REPORT REQUIREMENTS: There are two his-torical report requirements for proper documentation of argon-41 releases.

a. The SGM automatically prints out the amount of argon-41 released every 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Each of these 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> reports must be taped into the historical release data log.

i i

b. Finally, at the end of each day, as part of the re- l actor chut-down, an historical one hour report should ]

be printed and taped into the historical release data i log. This report contains the average counts per I minute for each of the 30 previous hours. This infor-mation can be used to calculate the amount of argo.7-41 released through through the reactor stack.

3. SYSTEM CHANGES OR OBSERVED ABNORMALITIES: Any changes to the system set points or observed abnormalities should be reported immediately to the Reactor Facility Director and to the Safety & Health Department.

I ,

I  !

l i

I l l

I

l 1 i Revised: Dec 1986 PROCEDURE IX REACTOR ROOM SAFETY l General: The following safety procedures will be observed while M the reactor room.

Specific:

1. Holst Operations: Perform the following before/during any hoist operations:

a. Inspect any lifting equipment (ropes, cables, etc.) for wear or damage prior to use.

b. Ensure that the hoist has a current load-testing (within last 12 months).

c. Ensure areas beneath the hoist are clear of personnel when operations are underway. This is particularly important when using the hatches between several floors.

d. Each time a load approaching 10,000 pounds is handled, test the brakes by raising the load a few inches, applying the brakes and checking for slippage.

e. Ensure a load is not lowered below the point where two full wraps of cable remain on the drum.

f. Ensure no tools or poles longer than 10 feet are raised vertically in .

the reactor room.

2. Mercury thermometers are not allowed in the reactor room at any time.

I l

I l

l I -

i

l' f i l

l ATTACHMENT C SAFETY ANALYSIS OF MODIFICATIONS TO UPGRADE THE REACTOR FACILITY AT THE ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE l

f i

1l l

l l

1

- _ _ _ _ . _ _ _ - _ _ - _ _ - _ _ _ _ _ _ _ l

l b.

I jl.

i-l f 10 CFR 50.59 SAFETY EVALUATION REPORT OF THE NEW REACTOR

-INSTRUMENTATION AND CONTROL SYSTEM AT THE ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE 1.

1 i

11 MAY 1988 Mark Moore Ken Hodadon Angela Munno I

i i

I i

l lJ i.

l l.

1

1 l

s l

ABSTRACT l This report describes changes to the reactor facility at the Armed Forces

! Radiobiology Research Institute (AFRRI) in Bethesda, Maryland. This Safety Evaluation Report (SER) meets the requirements of Title 10, Code of Federal Regulations, part 50.59 (10 CFR 50.59), and provides the basis for l

the conclusion that the changes to the facility involve no unreviewed safety questions and, in fact, are improvements in the facility design at AFRRI. In order to accomplish these changes, the Facility Safety Analysis '

Report (SAR) must be modified. The body of this report contains a description and safety analysis of the SAR changes. Excerpts from the SAR and the proposed changes are included as appendices.

Note: Under 10 CFR 50.59, a licensee may make changes to its facility provided that no changes are made to the Technical Specifications, and that there are no unreviewed safety questions. The conditions for unreviewed safety questions are outlined in 10 CFR 10.59.a.2, and are summarized below:

If the affected equipment is related to safety:

1. The probability of occurrence or the consequences of an accident or equipment malfunction shall not be increased.

ii. The possibility for an accident or malfunction of a different type than previously evaluated in the SAR shall not exist.

iii. The margin of safety as defined in the Basis for any Technical Specification shall not be reduced.

I i

l

TABLE OF CONTENTS u

I I. Abstract L II . - Table of Contents I

III. Introduction IV. Facility Modifications Safety Evaluation A. Overview l B. Reactor Safety System Descriptions

1. High Fin *; Safety Channels One and Two

2. Fuel Temperature Safety Channels One and Two

3. SCRAM Systems

4. Single Failure Criteria Analysis

5. TRIGA Reactor Safety System Failure Analysis C. Reactor Operational Instrumentation System Descriptions

1. Reactor Operational Channels

a. Multirange Linear Channel

b. Wide Range Log Channel

2. Reactor Interlocks (Rod Withdrawal Prevents)

3. Servo Controller

4. Rod Drives D. Reactor Modes of Operation E. Comparison of the Current and the New Reactor Safety and Control Systems

1. Reactor Safety Systems

2. Reactor Operational Control and Monitoring Systems 3.. Standard Control Rod Drives F. Safety Evaluation Conclusion APPENDICES: A. Listing of Corrections to be made to the SAR B. Proposed SAR Changes for the Previously Discussed Facility Modification Safety Analyses  !

C. AFRRI TRIGA Console (Safety) Scram System Single Failure Criteria Analysis D. Scram Circuit Safety Analysis for the University of Texas TRIGA Reactor E. Analysis of 5 Dollar Ramp Insertion Over a 2 Second Interval in AFRRI TRIGA Reactor

1 L

l INTRODUCTION Present conditions at the Armed Forces Radiobiology Research Institute

.(AFRRI) require that modifications be made to upgrade the reactor facility. The changes being made to the Facility Safety Analysis Report (SAR) include: The installation of a new Reactor Instrumentation and L Control System and the installation of three new stepping-motor standard

! control rod drives.

i AFRRI's current reactor instrumentation system is a 1972 vintage unit I f (hereafter, refered to as the current-(present), old, or 1972 console) salvaged from the 1977 decommissioning of the Diamond Ordnance Radiation i Facility and was' installed at AFRRI in 1978. The design life of this unit is 10 years. Because this console is now 16 years old, maintenance down time has increased and is expected to continue to increase over the next five years.  !

.The console's functional utility is now continuously diminishing due to the progressive obsolescence of many of its electronic components.

Although the obsolescence of these components does not effect the nuclear safety of-the. system, it is a problem operationally. Many of these electronic components are no longer manufactured; consequently, direct '

I replacements are unobtainable. Redesign of selected circuits to use currently available electronic components would require, in each case, a safety review by the reactor safety committee and possible review and approval by the NRC. i Ectimated hardware costs to entirely redesign, replace, and upgrade >

AFRRI's existing console exceed the cost of buying a new instrumentation system.

Failure analyses of current console components indicate that, under normal circumstances, AFRRI has sufficient spare parts to sustain its present operational capability for less than 2 years. Then it is expected that AFRRI-would become involved in serious down time problems.

1 AFRRI's control rod drive system also suffers from the same progressive 1 obsolescence, increasing maintenance down time, and spare parts I I

unavailability as the control console.

Acqu'eing a new state-of-the-art console and control rod drive system usin. .ntegrated circuits and microprocessor technology will resolve these probiems and provide for reliable operation of the AFRRI Reactor Facility through the year 2000. j l

.This new state-of-the-art microprocessor-based instrumentation and control system will replace the current control console while improving the i

existing operational capabilities and safety characteristics. The new system will increase reactor operational performance through increasec productivity, improved efficiency, increased reliability, improved

[

l l

l L

i l experiment reproducibility, and increased maintainability. productivity will be improved through increased reactor operating time due to the l cystem performing automatic self-checks of daily instrumentation I checkouts, and through decreased operator training time - operators will (

bacome proficient in a much shorter length of time. The new system will ;

increase efficiency in reactor operators' time by automatically logging l reactor data or allowing keyboard entry of nonoperational but essential l l

! information pertinent to reactor operations. Experiment reproducibility will be improved through increased pulse accura;y and repeatability and 1 i

through improved Auto Mode capabilities. In pulse Mode, the system will  !

l provide prompt waveform analysis: peak power, energy, half power width, l reactivity insertion, minimum period, and peak fuel temperature are l measured and calculated automatically and reported promptly to the In Automatic Mode, the {

j operator in either graphic or nongraphic mode.

operator will select the desired power level, run duration (SCRAM time),

and which rods will be servoed, then position the banked rods, select the j Automatic Mode and let the Reactor Control System perform the run. The -

new system will increase maintainability through state-of-the-art system maintenance design and layout, line replaceable units and on-line system diagnostics. System safety will also be improved through the performance of periodic self-diagnostics that determine if the unit is in a safe operational status. These diagnostics will display error messages reporting failures to the operator and will automatically place the reactor in a safe neutronic configuration. Additionally, the system will have improved Electromagnetic Interference (EMI) protection through shielding, optical isolation, and digitizing data at near core locations, and will reduce cabling requirements by collecting data in the reactor room and then routing that data to the Control Console computer via serial data trunks.

The Code of Federal Regulations (Title 10, part 50.59) requires that modification of a portion of a licensed facility as described in the fccility SAR be documented with a written safety evaluation. Such documentation provides the basis for determining that the change does not involve an unreviewed safety question. An unreviewed safety question according to 10 CFR 50.59 involves (1) the increase of probability of occurrence or the increase of consequences of an accident or malfunction of equipment important to safety compared to that situation previously evaluated in the SAR, or (2) the possibility for an accident or malfunction of a different type than previously analyzed in the SAR, or (3) the reduction in margin of safety as defined in the SAR.

Based on the analyses in this Technical Report, it has been determined that the proposed changes to the Reactor Facility do not involve any unreviewed safety questions and will actually improve the facility design at AFRRI.

This technical report describes changes and modifications made to the AFRRI reactor facility as depicted in the facility's SAR. These changes have been reviewed by the Reactor Facility Director and found to contain no unreviewed safety questions. This report is submitted to the Reactor

J L

f and Radiation Facility Safety Committee (RRFSC) for their concurrence that conditions of 10 CFR 50.59 are met. These conditions are that no

'unreviewed safe *.y questions are present and that the changes made do not increase tt.*: probability of occurrence or the consequences of an accident or malfunction.

l The proposed modifications require minor changes to the SAR. The body of l this report contains a description and safety analysis of the 10 CFR 50.59 SAR changes. Appendix A contains a specific page/section index of all of i

'the SAR changes. Appendix B contains excerpts from the SAR, for each of these 10 CFR 50.59 modifications.

[

The new Digital Reactor Instrumentation and Control System has been i designed to be safer than the present AFRRI control system which has been evaluated in the AFRRI TRIGA Mark F Reactor SAR. This has been accomplished by continuing to hardwire all safety circuits in a redundant, fail safe configuration. These safety circuits are completely independent of the data acquisition computer (DAC) and the control system computer (CSC). This means that if either or both computers were to fail, the failure cannot prevent the reactor from scramming.- On the other. hand, critical functions of the computers are monitored by " watch-dog-timers".

If the computers fail to update the timers in a predetermined fashion, the redundant, hardwired watch-dog-timers will scram the reactor.

1 As a result, the new Digital Reactor Instrumentation and Control System has equal or greater safety built-in than the present AFRRI control 1 system, which has SAR approval.

4 I

J l

l 1

l 1

i 1 \

L FACILITY MODIFICATIONS SAFETY EVALUATION l

i l' The installation-of_the new Reactor Instrumentation and Control System;at ]

tho AFRRI TRIGA Mark'F reactor facility will provide equal or greater

_ _ operational and safety capabilities with a higher degree of reliability than the current instrumentation.

OVERVIEW-L The basic elements of the new Reactor Instrumentation and Control System )

(see Figure 1) will consist of a Control Console,.a Data Acquisition and Control Unit (DAC), two independent power Monitor and Safety Systems, an Operational Channel, and a pulse Channel. This system was design and built in accordance with ANSI /ANS-15.15-1978 " Criteria For The Reactor

' Safety Systems of Research Reactors".

The. control Console will be a desk-type unit located in the AFRRI Reactor Control Room. Operators will conduct reactor operations using a set of control switches and a keyboard located on the console, r<nd the operators will receive feedback information through a high-resolution color monitor, a status monitor, indicators, and annunciators.

The heart of the control console will be the Control System Computer (CSC). Operators will adjust the rod positions by issuing commands-to the CSC,.which.will transmit these commands to the DAC. The DAC will reissue the commands to the drive' mechanisms. During reactor operations, the CSC will receive raw data from the DAC, process this data, and present the data in meaningful engineering units and graphic displays on a number of paripheral systems.

-The'CSC will operate two color CRT monitors. A high-resolution color graphics CRT (Reactor Control CRT) will provide the operator with a real-time graphic display of the reactor status. This CRT will display the important operational parameters using bar graphs and digital readouts and will alert the operator to any abnormal or dangerous conditions. A Reactor Status CRT will display pertinent diagnostic messages, reactor status, and~ facility status information.

The CSC will also interface with a near-letter-quality printer, allowing the logging of reactor information as required by the reactor operator.

Historical data will be saved in the CSC's internal memory and on command from.the operator be replayed, printed, or transferred to removable disks for permanent storage. This will provide the capability to maintain records of pertinent reactor statistics and to replay reactor operational records for training and analysis. In addition, the CSC will operate a color graphics ~ printer capable of printing steady-state and pulse mode data as well as producing point-line plots. Finally, the CSC will

I I interface to real-time recorders of reactor power and fuel temperature .

The DAC will be located in the AFRRI Reactor Room adjacent to the reactor and will provide high-speed data acquisition and control capability. The DAC will monitor the two independent power Monitor and Safety Systems, the Operational and Channel, temperature, and the pulse control rodChannel, the fuel temperature, water level positions.

the CSC, reissue the commands to The DAC will, on command from the reactor. raise and lower the control rods or scram trunks. The DAC will communicate with the CSC via serial data trunk fail. The These secondary trunk will serve as a backup should the primary serial data trunks will drastically reduce requirements between the Reactor Room and the Control Console. the wiring The power Monitor and Safety Systems will monitor the power from 1% to 120%

the of full event of power (1.0 megawatts) an overpower and shut the reactor down (SCRAM) in condition. The Operational Channel will monitor the power from source level to full power and the rate of power change (from -30 to +3 second period) in the steady state modes.

The pulse pulse mode.Channel will monitor the power level up to 5000 megawatts in the This channel will use an ion chamber, or some other acceptable pulse monitoring detector. a photo diode detector, information from the pulse channel and transmit The DAC will collect processing. the data to the CSC for The control console will have 8 Hardwired (Analog) LED Bargraph indicators which are located on the left side of the console. These hardwired channels include the two High Flux Safety Channels, the two Fuel Temperature Safety Channels, the Operational Wide-Range Log Channel, the period Channel, and the pulse NV and NVT Channels. Located below these analog bargraphs Temperature are the Channel stripOperational Multirange Linear Channel and Fuel chart recorders.

and are completely independent of the CSC and These therefore, will provide items are all hardwired DAC computers, and even should the CSC and DAC computers fail.information to the reactor operator at all times AFRRI Drives is also replacing its three 1960 vintage Standard Control Rod systems. with three new Standard Control Rod Drives using pulsed motor drive These stepping notars operate on phase-switched de power. These motors drive a pinion gear (connected to the Magnet Draw Tube) and a 10-turn positive feedback potentiometer via a chain and pulley gear mechanism. Except for the drive motors, the new control rod drive i assemblies will be the same as the current control rod drive assemblies.

REACTOR SAFETY SYSTEM DESCRIPTIONS HIGH FLUX SAFETY CHANNELS ONE AND TWO l

i l

i

!iI  ! jl<

s

- 0 ,

F m

.' 0 r

2

.,. 0 u K M 0 I R E M e

1

1. A T A

.,

• 0 M S R E Y G 0 m b9 l l

G A S A I u I L D 0 . G R O '

r s

u I

F T RT KC n

nA I

R N O R O L Il F

A C D n . s n

a I

, l

.a f km s t n

en es a

s r ,

a, a

c ,,,

r Il n

A e a a - m

~

s _ o s

M O .s"m e

m I

E 8

e m

m s

e m m m s a

O m8"- e e

. o a M C n c se y Il L + . s s

s a ,

O 4 a m h s*,

e er se e r e

C a c m", m a mse n ,

c ee o d c.*a*"" s 8

i " 3 .

• ' C m a r

a*3 pn i

r emS Q

0 C

.e e

D G

m e a

T p s

T w

r I[

wm r

e m

e e de fl r

i a

P C=

T s s s s e .a m" m n sa "

m _

F _

_ u e i ear8' e

,. * 'a e. L e ."ha

. e n F e

_T"I n em e

• o s m k I i!i j s r a nr eul l ,E l I e

s m m

d. e e

y e nl l li. I I I l I g 1 r,I l !

l il as am e m-ne su m e T a L

L t_ m T e

A . n r H na e e

J t ea nd sp u s

r u

s l 8 r

u. s C .

oemMA m .co r e I . m a 3 tlt m a =e eee m m

C Ac _ = nes m A ,

_ in

== eeA a E

R ug _-

=

D r

e =

=

e r c r

e c )

m m jsl m

. . e o m

. ) = r c, e a s

s

.r e

s e

e T

(

T m

. = r u

e .

e.

s r

=

- r e

s

= = _

.F( . dn a s e r

. e m

.m r

e s

e e

s eWA e

e s

r o

=

=

T e i _

= 4 w l r c _

_=

= = I F"

r _

i m_ n ,= I r =

. e r e

. n m ese ea e

sn c ,

, h C  %=a=

• a c

1 2 1 2 C

1. 8 I 5 O y L t p

a iuu s

p. xx ce af e el l ea TTFF B s 1 1 hh eegg u uii F FHH

lI l

High flux safety channels one and two report the reactor power level as l

[ measured by two ion chambers and a pulse detector placed above the core in j the neutron field. Each safety channel is a part of one multifunction l l NP-1000 neutron power channel. For safety reasons (simple redundancy) two l independent NP-1000's are used and they operate identically during steady i state operation. Each channel consists of an ion chamber placed above the f

,I I

core and the associated NP-1000 electronics. The steady state power level is displayed on two separate LED bargraph indicators and on the reactor l

I control CRT.

During pulse operation, high flux safety channel one is shunted and the sensor for high flux safety channel two is switched to a third,

, independent pulse detector placed above the core. High flux safety )

channel two measures the peak power level achieved during the pulse (NV) J and the total integrated power produced by the pulse (NVT) and is therefore specified as an NPP-1000 instead of an NP-1000. However, it I should be noted that both safety channels operate with identical NP-1000 circuitry. Calibration of the NP-1000's is done automatically during the Daily Startup Checklist when the operator initiates the " pre-checks" by i

I activation of the Prestart Check Switch on the control console's Mode l Control Panel. Any failures detected during the prechecks will be I automatically reported to the operator via the reactor status CRT.

The high flux safety channels (NP-1000's) form part of the scram logic i circuitry. When the steady state reactor power level, as measured by j either high flux safety channel, reaches the maximum power level specified -

in the technical specifications, a bistable trip circuit is activated which breaks the scram logic circuit, causing an immediate reactor scram. ,

Similarly, when the reactor power level during pulse operation, as  !

measured by high flux safety channel two, reaches the maximum pulse power I level specified in the technical specifications, a bistable trip circuit is activated which causes an immediate reactor scram.

FUEL TEMPERATURE SAFETY CHANNELS ONE AND TWO Fuel temperature safety channels one and two are independent of one another but operate in identical manners (simple redundancy). One thermocouple from each of the two instrumented fuel elements, one in the l B-ring and one in the C-ring, provide inputs to fuel temperature safety i channels one and two, respectively. The two fuel temperature signals are amplified and displayed on two separate bargraph indicators located on the reactor console and on the reactor control CRT. The fuel temperature safety channels have internal compensation for the chromel-alumel I thermocouple and high noise rejection. Calibration of the Fuel Temperature Channels is done automatically during the Daily Startup Checklist when the reactor operator initiates the " pre-checks" by activation of the Prestart Check Switch on the control console's Mode Control Panel. Any failures detected during the prechecks will be automatically reported to the operator via the reactor status CRT.

l l

l

l l

In addition to providing information to the reactor operator on fuel temperature, the fuel temperature safety channels also form part of the ceram logic circuitry. When the fuel temperature, as measured by either

! fuel temperature safety channel, reaches the maximum allowable fuel l temperature specified in the technical specifications, a bistable trip circuit is activated which breaks the scram logic circuit, causing an inmediate reactor scram. The operational fuel temperature limit is l

usually set below the technical specifications limit to assure an adequate i dogree of reactor protection.

The combination of the two independent High Flux Safety Channels and the two independent Fuel Temperature Safety Channels provides both simple redundancy and functional redundancy in terms of insuring that the Reactor l Safety Limit as specified in the Technical Specifications is never i reached.

SCRAM SYSTEMS i

The scram logic circuitry (see Figure 2) assures that a set of reactor l core and operational conditions must be satisfied for reactor operation to '

occur or continue in accordance with the technical specifications. The ceram logic circuitry involves a set of open-on-failure logic relay switches in series: any scram signal or component failure in the scraa logic, therefore, results in a loss of standard control rod magnet current end a loss of air to the 'ransient rod cylinder, resulting in a reactor scram. The time between e.ctivation of the scram logic and the total insertion of the control rods is limited by the technical specifications to assure the safety of the reactor and the fuel elements for the range of anticipated transients for the AFRRI TRIGA reactor. The scram logic circuitry causes an automatic reactor scram under the following circumstances:

- The steady state timer causes a reactor scram after a given elapsed time, as set on the timer, when utilized during steady state power operations.  ;

- The pulse timer causes a reactor scram after a given elapsed time, as set on the timer (in accordance with the limit specified in the technical specifications), during pulse power operations.

- The manual scram button located on the reactor console, allows the Reactor Operator to manually scram the reactor.

- Movement of the console key to the OFF position causes a reactor scram.

- The reactor tank shielding doors in any position other than fully open or fully closed will cause a reactor scram (this is part of the facility interlock system).

1

m e m m a r a a r r r n r e c e e u r m ire r S t s c c w e o S S i e e o f ed p T m s d v P t e px i o e es g g e T m R oE o t a l

C y sol t d

o a e r l r A e Sd . d t d c a e sC n h h S o S K r ya c c u t f ed c M d n e on t t y l i u o l Ba ne a a d e a v el W W a e s s o go e u i l e ds e rs l

s a

n d e s m eep ee C C t t e e u o o ee S A S F g f D

{L {C N{0 BC_ _C -

S M

A H

C S

R O

m( '

T C

A -

E f R

t r m o t e i F a e e A l I r s n r p R c e g t d p R S m e s n

F .

I g

I O a h A

2 ,' ,* ) ) ) )-

E R

U G

I F

7 ]1 7j 2 7d 3J 1

t e

l e

2 n

_ '1 .~ - # #

m m n m 2 m y y Va V e

1. u # u Hs N8 :

_ l l m

e m

e f O f 0

y1F y1F T 0 oy t 0 t0 T *0 oy e1 e n t t

_ l e1 i l 0 l0 se s e f1 e f1 g e6 e6 ef of

_ e i a s u

_ S N S i

N t

F F ma I s pSa i

i:

l' I '

- Activation of any of the emergency stop buttons in either exposure room or on the console causes a reactor scram.

- A loss of AC power to the reactor causes a reactor scram.

- High flux safety channel one causes a reactor scram at a reactor power

level specified in the technical specifications for steady state modes i of operation. This may be operationally set more conservative than the technical specifications limit. l l

- High flux safety channel two causes a reactor scram at a reactor power i level specified in the technical specifications for steady state modes of operation. This may be operationally set more conservative than the 4 technical specifications limit. j

-A loss of high voltage to either of the detectors for high flux safety channels one and two causes a reactor scram.

- Fuel temperature safety channels one and two will each initiate a reactor scram if the fuel temperature, as measured independently by either channel, reaches 600*C (technical specification limit). This )

assures that the AFRRI safety limit (core temperature) of 1,000*C for l AFRRI stainless steel clad cylindrical TRIGA fuel elements, as stated in the AFRRI technical specifications, is never approached or exceeded. I The actual operational limit for the fuel temperature safety channels may be set lower than the technical specifications limit of 600*C.

- A loss of reactor pool water which leaves less than or equal to 14 feet of pool water above the core (technical specifications limit) causes a reactor scram. The actual operational limits for the pool water level may be set more conservatively than the technical specifications limit.

- One watchdog timer on the data acquisition computer and another one on the control system computer are required to be reset periodically by a program routine as a safeguard against computer component failures either in hardware or software. If the required response is not received within a definite time period, redundant normally open (fail safe) contacts interrupt the scram loop dropping the rods and shutting down the reactor. These watchdog timers are additional safety devices.

SINGLE FAILURE CRITERIA ANALYSIS ANSI /ANS STD 15.15-1978 " Criteria for Reactor Safety Systems of Research Reactors" specifies that a Single Failure Criteria Analysis be performed on all non-redundant reactor safety systems. This analysis was performed by General Atomics for the new AFRRI TRIGA Reactor Instrumentation and Control System and is enclosed as Appendix C "AFRRI TRIGA Console (Safety)

Scram System Single Failure Criteria Analysis." This analysis

l 11 demonstrates that, except for the Reactor Key Switch (which does not perform a safety function except to prevent unauthorized startup), the Mean Time Between Failure of any single element of the new instrumentation scram system greatly exceeds (the MTBF's range from 23 years to 125 years)'

l the design life of the new coneole (15 years). This analysis was performed for any single failure of the reactor safety system.

ll TRIGA REACTOR SAFETY SYSTEM FAILURE ANALYSIS Although not required, a Failure Analysis was performed by the University of Texas and General Atomics of the new Reactor Instrumentation and Control System. This analysis is enclosed as Appendix D "TRIGA - ICS Reactor Safety System Failure Analysis". This analysis looked at the probability of the Reactor Safety System failing to perform its intended function: no scram occurs during a scram situation. In order for this to occur there would need to be simultaneous failures of two or more components of the Reactor Safety System. This analysis demonstrates that the Probability of Failure of the new Reactor Safety System is 2X10- 21 failures / hour, or a mean time between failures of SX108 years.

REACTOR OPERATIONAL INSTRUMENTATION SYSTEM DESCRIPTIONS REACTOR OPERATIONAL CHANNELS Multirange Linear Channel I The mulitrange linear channel is one of three channels included in the NM-1000.

I The multirange linear channel reports reactor power from source level

[ 8 Wt (thermal watts)) to full steady state power (1 MWt).

of a principle fission detector serves as the channel input.

The output The channel consists of two circuit sections: the count rate circuit, and the campbelling circuit. At power levels less than 1 kilowatt (t) the count rate circuit is utilized. The count rate circuit generates an output voltage proportional to the number of neutron genrated pulses or counts I received from the fission detector. Hence, the output is proportional to the neutron population and the reactor power level. For steady state power levels at or above 1 kilowatt (t) the campbelling circuit is I utilized. The campbelling circuit generates an output voltage proportional to the reactor power level by a verified technique of noise envelope amplitude detection and measurement known as campbelling. The NM-1000's micro-processor converts the signal from these circuits into 10 I linear power ranges. This feature provides for a more precise reading of linear power level over the entire range of reactor power.

i l The NM-1000's multirange linear channel output is displayed in two formats. These are a bargraph indicator on the Reactor Control CRT 5

display and a strip chart recorder located on the left-hand vertical panel I on the control console. As a performance check, the microprocessor )

automatically tests the channel for campbell circuit operability while the I reactor is operating in the count rate range and vice verse when the reactor is in the campbelling range. The multirange ranging function is i auto-ranged via the NM-1000 control system computer, j l

Wide Range Log Channel j The wide-range log channel like the multirange linear measures reactor power from source level ( 8 Wt) to full steady state power (1MWt). It is a digital version of the General Atomics 10-decade log power system to cover the reactor power range and provide a period signal. For the log i power function, the chamber signal from startup (pulse counting) range i through the campbelling [ root mean square (RMS) signal processing] range covers in excess of 10-decades of power level. The self-contained microprocessor combines these signals and derives the power rate of change  ;

(period) through the full range of power.

The wide-range log channel forms part of the rod withdrawal prevent (RWp) interlock system. The channel activates variable set point bistable trips in the rod withdrawal prevent interlock system if source level neutrons ,

( 8 Wt) are not present, if the reactor power level is above 1 KWt I when switched to pulse mode, if a steady state power increase has a period of 3 seconds or faster during certain steady state modes, or if high i voltage is not supplied to the fission detector.

The wide-range log and period output are displayed on bargraph indicators which are both hardwired and on the Reactor Control CRT. The NM-1000's microprocessor, similar to the multirange linear channel, automatically j tests the wide-range log channel for upper and lower decade operability. j I REACTOR INTERLOCKS (ROD WITHDRAWAL PREVENTS)

A Rod Withdrawal prevent (RWP) interlock stops any upward motion of the standard control rods and prevents air from being supplied to the transient control rod unless specified operating conditions are met. An RWp interlock, however, does not prevent a control rod from being lowered or scrammed. Therefore, any RWp interlock prevents any further positive reactivity from being inserted into the core until specific conditions are satisfied.

I The system of RWp interlocks prevents control rod withdrawals under the following circumstances: q l

- RWp prevents air from being applied to the transient rod unless the )

reactor power level is under 1 KWt. l I i i

1 i

l L

l

( - RWP prevents any control rod withdrawal unless, as a minimum, source l level neutrons ( 3 Wt) are present.

' - RWP prevents any further control rod withdrawal unless the power level is changing on a 3-second or longer period as measured by the wide-range log channel during certain steady state operations, f

l .RWP prevents any control rod withdrawat unless high voltage is being supplied to the fission detector for the multirange linear and wide-range log channels.

l RWP prevents any control rod withdrawal unless the bulk pool water temperature is less than 60*C (Technical Specification Limit).

SERVO CONTROLLER The Servo Controller, in the Automatic and Square Wave Modes, controls the reactor power automatically to within +/-1% of the demand power level selected by the operator. Thumbwheel switches are provided on the Mode Control panel for the desired power selection. The Servo Controller will track and stabilize reactor power through the utilization of a PID algorithm (Proportional, Integral, Derivative). The console will be capable of servoing any combination of the three standard control roda (REG, SAFE, or SHIM). It will not, however, servo the Transient Rod in any mode. The operator will be able to select which combination of rods will be servoed via a Servoed Rod Selector Switch located on the Mode Control Panel of the new control console. The Servo Controller system utilizes the latest digital computer technology coupled with extensively developed software. The current console uses an analog computer to servo the rods while the new console uses a digital computer to servo the rods.

Reactor flux level and change is accurately and rapidly measured by an analog / digital input from the Operational (fission) Channel. The PID algorithm in the DAC then responds to this input as compared to the operator set Demand Power Level Setting through the servoed control rods I

which are powered by precise translator / stepping motor drives. The i (operator selected) drive (s) will be driven up or down automatically to control the power level to within +/-1% of the Demand Power Level Setting.

The new console Servo Controller can drive all three standard control rods I

simultaneously (- $ 5. 50) in the Automatic and Square Wave Modes versus the old console which can servo the Transient and the REG rods (- $ 5. 50 )

simultaneously in the Square Wave Mode and which servoed the REG rod in the Automatic Mode; by technical specifications the maximum excess reactivity above cold critical is $5.00. A Ramp Accident Analysis was performed to insure that a runaway drive situation involving a two second full-insertion (this is faster than the maximum drive rate of the new drives) of all three standard control rod drives would not lead to an I

i l

t event. This analysis was performed by General Atomics under contract to AFRRI and is enclosed as Appendix E " Analysis of a Five Dollar Ramp

! Incertion Over a Two Second Interval in AFRRI TRIGA Reactor". This

! analysis demonstrates that the consequences of this tecident scenario are trivial. The peak power level attained is 330MW and the maximum fuel temperature attained is 330'C. The AFRRI TRIGA Reactor routinely pulses to peak powers of up to 3300MW and the normal 1 MW steady state fuel temperature is approximately 420*C. This analysis demonstrates that there are no unreviewed safety questions.

ROD DRIVES The rod drive mechanisms for each of the new Standard Control Rod Drives is an electric stepping-motor-actuated linear drive equipped with a magneti; coupler and a positive feedback potentiometer. The purpose of each of the rod drive mechanisms is to position the reactor control rod elements.

General Operational Description A stepping motor drives a pinion gear and a 10-turn potentiometer via a chein and pulley gear mechanism. The potentiometer is used to provide rod pocition information. The pinion gear engages a rack attached to the mognet draw tube. An electromagnet, attached to the lower end of the draw tube, engages an iron armature. The armature is screwed and pinned into the upper end of a connecting rod that terminates at its lower end in the control rod.

When the stepping motor is energized (via the rod control Up/DOWN switch on the operator's console), the pinion gear shaft rotates, thus raising the magnet draw tube. If the electromagnet is energized, the armature and the connecting rod will raise with the draw tube so that the control rod is withdrawn from the reactor core. In the event of a reactor scram, the magnet is de-energized and the armature will be released. The connecting rod, the piston, and the control rod will then drop, thus reinserting the control rod into the core.

Stepping motors operate on phase-switched de power. The motor shaft advances 200 steps per revolution (1.8 des per step). Since current is maintained on the motor windings when the motor is not being stepped, a high holding torque is maintained.

The torque vs speed characteristic of a stepping motor is greatly dependent on the drive circuit used to step the motor. To optimize the torque characteristic vs motor frame size, a Translator Module was selected to drive the stepping motor. This combination of stepping motor and translator module produces the optimum torque at the operating speeds of the control rod drives.

1:

r l

l REACTOR MODES OF OPERATION Thore,are four standard operating modes: manual, automatic, square wave, l

'. and pulse.

l The manual and automatic modes apply to the steady-state reactor condition; the square-wave and pulse modes are the conditions implied by their names and require a transient (pulse) rod drive.

The manual and automatic reactor control modes are used for reactor operation from source level to 100% power. These two modes are used for manual reactor start up, change in power level, and steady-state operation. The square-wave operation allows the power level to be raised quickly to a desired power level. The pulse mode generates high-power levels for very short periods of time.

Manual rod control is accomplished through the use of push-buttons on the rod control panel. The top row of push-buttons (magnet) is used to i interrupt the current to the rod drive magnets. If the rod is scrammed.

and the drive is above the down limit, the rod will fall back into the core and the magnet will automatically drive to the down limit, where it again contacts the armature.

The middle row of push-buttons (up) and the bottom row (down) are used to pocition the control rods. Depressing these push-buttons causes the control rods to move in the direction indicated. Several interlocks provent the movement of the rods in the up direction under conditions such as the'following:

1. Scrama not reset.

2. Magnet not coupled to armature.

3. Source level below minimum count.

4. Two UP switches depressed at the same time.

5. Mode switch in the pulse position.

6. Mode switch in automatic position (servoed rods only).

7. period less than 3 seconds.

There is no interlock inhibiting the DOWN direction of the control rods except in the case of the servoed rods while in the AUTOHATIC mode. In all cases, however, the manual scram of any rod will result in the full insertion of the rod into the core.

Automatic (servo) power control can be obtained by switching from manual operation to automatic operation via operator activation of the Auto Mode Switch on the control console's Mode Control panel. All the instrumentation, safety, and interlock circuitry described above applies and is in operation in this mode. However, the selected servoed rods are now controlled automatically in response to a power level and period signal. The reactor power level is compared with the demand level set by

l l

the operator and is used to bring the reactor power to the demand level on e fixed preset period. The purpose of this feature is to automatically mnintain the preset power level during long-term power runs. Options are available to the operator to maintain power by movement of a single rod or by bank operation of selected rods. The rods to be servoed are selected by the operator via the Servoed Rod Selector Switch on the control console's Mode Control panel.

In a square-wave operation, the reactor is first brought to a critical condition below 1 KW, leaving the transient rod partially in the core.

All of the steady-state instrumentation is in operation. The transient rod is ejected from the core by means of the transient rod FIRE push-button. When the power level reaches the demand level, it is maintained in the same manner as in the automatic mode.

Reactor control in the pulsing mode consists of establishing criticality at a flux level below 1 KW in the steady-state mode. This is accomplished by the use of the motor-driven control rods, leaving the transient rod either fully or partially inserted. The mode selector switch is then depressed. The Transient Rod Fire switch automatically connects the pulsing chamber to monitor and record peak flux (nv) and energy release (nyt). pulsing can be initiated from either the critical or suberitical reactor state.

COMPARISON OF THE CURRENT AND THE NEW REACTOR SAFETY AND CONTROL SYSTEMS REACTOR SAFETY SYSTEMS The current console, which was designed and built in the early 1970's, has as its Reactor Safety Systems (See Table I) two hardwired independent ,

analog High Flux Safety Channels, two hardwired independent analog Fuel l Temperature Safety Channels, and a hardwired relay logic SCRAM circuitry. )

The High Flux safety Channels derive their signals from two Boron j (neutron sensitive) Ion Chambers mounted above the core, and these channels have readouts located on the vertical panel of the control console in the form of analog meters. The Fuel Temperature Safety Channels derive their signals from two instrumented fuel elements, one located in the B-ring and one located in the C-ring. The Fuel Temperature Safety Channels also have readouts located on the vertical panel of the control console in the form of analog meters. The Scram circuitry has two independent relay contacts for each safety channel, one located in the supply side and one located in the return side of the magnet and solenoid power circuitry. Dropping any one of these numerous relays would cut power to the magnets and the sir solenoid.

The new console, as with the old console, also has as its Reactor Safety Systems two independent hardwired analog High Flux Safety Channels, two i independent hardwired analog Fuel Temperature Safety Channels, and a j l

l J

i

)

I hardwired relay logic SCRAM circuitry. The High Flux Safety Channels, Just like the old console, derive their signals from two Ion Chambers

)i mounted above the core and have readouts located on the vertical panel of the control console. However, for the new console, these readouts take the form of LED bargraphs instead of meters. These new channels were designed to be the same as the old channels, only updated with current technology electronics. The Fuel Temperature Safety Channels will still derive their signals from the same two instrumented fuel elements located in the B-ring and in the C-ring. As with the High Flux Channels, the Fuel Temperature Channels have their readouts on the control console in the form of LED bargraphs instead of meters. It should be emphasized again, that these safety systems on the new consoles are independent hardwired analog channels just as those are on the old console. These systems are I completely independent of the system's computers and will continue to function irregardless of the state these computers are in. This will j

j insure safety system monitoring and control at all times. The Scram circuitry, again as with the old console, has two independent relays for each safety channel, one located in the supply side and one located in the I return side of the magnet and solenoid power circuitry. Similar to the four safety channels, the Scram circuitry was designed to be the same as {

the old Scram circuitry only replaced with current technology electronics. {'

Table 1 shows a comparison between the SCRAMS on the new and old consoles.

The SCRAM circuitry on both systems is the same except for the Safety Channel Calibrate Scram on the old console and the Watchdog Scrams on the new console. The old console used to shunt the inputs to the safety i channels while putting in calibration signals to the safety channels. j This created the possibility of operating with a safety channel in the calibrate mode. To prevent this condition from occurring the old console had a relay which would scram the reactor if any of the safety channels were switched to the calibrate mode. In the new system, the calibration signals are additive to the normal safety channel signals (e.g. the safety channels are not shunted in the calibration mode). A calibration signal added to the normal safety channel signal is more conservative (will always provide a higher channel reading) and therefore does not require a calibrate scram. However, watchdog scrams, as described earlier, have been added to the new console scram circuitry. These watchdogs monitor the status of the DAC and CSC computers and should any of the four watchdogs (two in the DAC and two in the CSC) fail to be reset by the software, then the system would scram the reactor. This ensures that failure of either of these computers or of their software will cause a I system scram.

I REACTOR OPERATIONAL CONTROL AND MONITORrNG SYSTEMS The 1972 console has an operational channel which derives its signal from a fission chamber and generates the Wide-Range Log and Multirange Linear monitoring channels. The operational channel combines the standard techniques of Count Rate and Campbelling in an analog computer to provide the capability to monitor 10 decades of power. The new console uses an

L

[

l ANALOG (1972) vs DIGITAL (1988) CONTROL CONSOLES ,

OLD ,

l SAFETY INTERLOCKS CONTROL DRIVES SYSTEMS (OPS CHANNEL)

Hardwired Relay Analog Phase Amp-8T Logic Computer Interrupt ,

circuit  ;

NEW .

SAFET( INTERLOCKS CONTROL DRIVES SYSTEMS (OPS CHANNEL)

Hardwired Firmware Digital Stepping Amp-BT NM-1000 Computer Motor circuit Relays & (Digital)

EPROM

CONSOLE REACTOR SAFELY SYSTEM COMPARISON l

OLD NEW  !

SAFETY CHANNELS SAFETY CHANNELS 1

- 2 Percent Power - 2 Percent Power l - 2 Fuel Temperature - 2 Fuel Temperature SCRAMS SCRAMS 4

-TECH SPEC -TECH SPEC

- 4 High Level Safety - 4 High Level Safety Trips Trips

- Manual - Manual

- 2 HV Loss % Power - 2 HV Loss % Power 1

- Pulse Timer - Pulse Timer  !

- Emergency Stop - Emergency Stop  !

- Water Level - Water Level j

-SAR -SAR i

- Key Switch - Key Switch

- Steady State 11mer - Stecc) State Timer ,

- Loss of AC - Loss of AC

- Facility Interiocks - Facility Interiocks

• Safety Channel Calibrate

• Watchdog

-- 2 Reicys in both the DAC and the CSC

- Individual Rod SCRAM - Individual Rod SCRAM i

L p

t

! CONSOLE INTERLOCKS COMPARISON 1

l OLD NEW

-TECH SPEC -TECH SPEC

- 1 kw - 1 kw

- Source Level Neutrons - Source Level Neutrons

- Mode I (no two rods) - Mode I (no two rods)

- Mode ill - Mode 111 (no rod except TRANS) (no rod except TRANS)

-SAR -SAR ,

i

- 3 second period - 3 second period j

- Ops Channel HV loss - Ops Channel HV loss I

- Bulk Water 60 C - Bulk Water 60 C

• 0ps Channel Calibrate

• (calibrate signal additive)

)

i

i f

operational channel which was designed to be a digital version of the old cyatem; it still combines the standard techniques of Count Rate and Cc=pbelling to provide the capability to monitor 10 decades of power. .

l Tho difference is that this function is now performed with a digital computer instead of a analog computer and uses current technology { ,

electronics. These two systems were demonstrated to be essentially l

equivalent during the manufactures test program when both the old and the I new systems were operated in parallel.

The interlocks or Rod Withdrawal prevents (RWps) for both the new and old systems are shown in Table 2. Again, these interlocks are the same for '

both systems except for the Operational Channel Calibrate RWp oc. the old console. On the old console, the input signal to the operational channel would be shunted when the channel was placed in the calibrate mode. In order to prevent operation of the reactor in this configuration, an RWp was added to the system to prevent rod withdrawal with the operational channel in the calibrate mode. On the new console, the calibration signal is additive to the normal operational signal, and again is therefore more conservative and requires no RWP. The interlocks on the old console were all analog logic asing relays. The interlocks on the new console use Digital Logic (Firmware).

STANDARD CONTROL ROD DRIVES The three standard control rod drives will be replaced. The old drives used phase-interrupt (analog) motors while the new drives will use stepping (digital) motors (See Table 3). Only the drive motors are being changed, the remainder of the control rod drive assemblies will stay the same.

SAFETY EVALUATION CONCLUSION The AFRRI TRIGA Reactor, NRC Facility License No. R-84. is classified as a

" Negligible Risk Research Reactor (Pulsing)" in accordence with the NRC approved AFRRI TRIGA Reactor Facility Safety Analysis and es defined in ANSI /ANS 15.15-1978 " Criteria for the Reactor Safety Systems of Research ,

Reactors". A " Negligible Risk Research Reactor (Pulsing)", as defined in {

ANSI /ANS 15.15-1978, is "a research reactor for which, in the postulated event of the complete failure of the reactor safety system coincident with the occurrence of the most adverse Design Basis Event, the radiological i consequences would be negligible." Pulsing is defined as "a reactor that '

has been specially designed with an inherent shutdown mechanism sufficient to allow the reactor to accept large reactivity insertiens without exceeding any safety limit." I I

In analyzing the safety of the AFRRI TRIGA Peactor, it is important to l ctart with the inherent safety of the TRIGA Fuel, which is designed to q l

l l

f operate with large positive step reactivity insertions. The inherent safety of the fuel element stems from its large prompt negative temperature coefficient of reactivity, which causes the automatic

! termination of a power excursion before any core damage results. The j Prompt Negative Coefficient of Reactivity of the AFRRI TRIGA Reactor is 0.0126 %deltaK/K per 'C (-1.7 cents /*C), while the Steady State Negative l

Coefficient of Reactivity is - 0.0051 %deltaK/K per *C ( .7 cents /*C).

! Fuel elements with 8.45 wt.%U have been pulsed repeatedly in General Atomics' Advanced TRIGA prototype Reactor (ATpR) to peak power levels of over levels greater than 2,000 MW.

8,000 MW, and have been pulsed thousands of times to peak power The AFRRI TRIGA Reactor is limited to a

$4.00 step positive reactivity insertion (technical specification limit) which would yield a peak power level of approximately 4,700 MW.

The AFRRI Facility Safety Analysis Report has analyzed two Design Basis Accidents.

The first Design Basis Accident, called the " Fuel Element Drop Accident," involved the postulated occurrence of a cladding failure of a fuel element after a 2-week period where the saturated fission product inventory of a 1 MW steady state operation has been allowed to decay after being taken out of the operating core and placed in storage; the saturated fission product inventory is obtained after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of continuous reactor operation at full power (1 MW). The cladding failure could occur when the fuel element is withdrawn from the reactor pool. While the fuel element is exposed to air, a cladding failure could occur coincidentally, or due to a drop. As the AFRRI FSAR explains, the probability of such an accident is considered to be extremely remote. The second Design Basis Accident, called the Fuel Element Cladding Failure Accident, involved the postulated occurrence of a cladding failure of a fuel element during a pulse operation or inadvertent transient following a steady state operation of 1 MW. Again, it was assumed a saturated fission product inventory which occurs after 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> of continuous reactor operation at full power (1 MW), and a pulse operation with an integrated energy of 40 MW-sec. A 40 MW-sec pulse operation is roughly *quivalent to a step positive reactivity insertion of approximately $4.50. The maximum worth of the AFRRI TRIGA Pulse Rod (Transient Rod) is approximately $3.75, and as such a 40 MW-sec pulse operation is an extremely conservative assumption. The AFRRI FSAR again explains that the probability of such an accident is considered to be extremely remote.

The analysis in the AFRRI FSAR shows that "... the consequences from the i

{

Design Basis Accident of a fuel element drop accident or a fuel element I clad failure accident were insignificant." Therefore, it was

"... concluded that the operation of the AFRRI reactor in the manner f

{

authorized by Facility License No. R-84 does not represent an undue risk j to the health and safety of the operational personnel or the general public."

Both of these Design Basis Accidents (DBAs) were postulated on the occurrence of one or two predetermined, deliberate man-made events. In the first DBA, the scenario required that the reactor be operated l

l l

continuously for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> at full power to build up a saturated fission prod;ct inventory. In the second DBA, the scenario again requires a saturated fission product inventory followed by a step positive insertion I of reactivity that produces 40 MW-see of integrated energy. AFRRI has never operated at full power for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> continuously, nor will probe.bly ever operate in this manner under normal operating conditions. Both of I these DBAs require fuel cladding failures following a set of specific man-made conditions and are not a result of any failures on the part of the Reactor Safety Systems. It was shown previously that the new console has a MTBF of the Reactor Safety System of 5 X 108 years. Failure of the Reactor Safety System would not initiate a Design Basis Accident. Even should the Reactor Safety System suffer a complete failure at the same moment as a DBA, the consequences would be negligible.

It was determined during the design of the new Reactor Instrumentation and >

Control System that no technical specification changes would be required.

There are no technical specification changes associated with the installation or operation of AFFRI's new Reactor Instrumentation and Control System.

I The new Reactor Instrumentation and Control System will offer a dramatic improvement in operational productivity, system reliability, and system maintainability.

The new Digital Reactor Instrumentation and Control System has been designed to be safer than the present AFRRI control system. This has been accomplished by continuing to hardwire all safety circuits in a redundant, I fail safe configuration. These safety circuits are completely independent of the data acquisition computer (DAC) and the control system computer (CSC). This means that if either or both computers were to fail, the failure cannot prevent the reactor from scramming. On the other hand, critical functions of the computers are monitored by " watch-dog-timers".

If the computers fail to update the timers in a predetermined fashion, the redundant, hardwired watch-dog-timers will scram the reactor. As a result, the new Digital Reactor Instrumentation and Control System has equal or greater safety built-in than the present AFRRI control system, which has SAR approval.

Based on the analyses in this technical report, it has been determined that the proposed changes to the Reactor Facility do not involve I unreviewed safety questions and, in fact, are improvements in the facility design at AFRRI.

I I

I I

l l

l l

1 This technical report describes changes and modifications made to the '

AFRRI reactor facility as depicted in the facility's SAR. These changes have been reviewed by the Reactor Facility Director and found to contain I no unreviewed safety questions. This report is submitted to the Reactor and Radiation Facility Safety Committee (RRFSC) for their concurrence that conditions of 10 CFR 50.59 are met. These conditions are that no I i

unreviewed safety questions are present and that the changes made do not increase the probability of occurrence or the consequences of an accident or malfunction.

1 I

)

i I

i l

I  !

I 1 I

I l

l I - _._ - _ - _ _ _ _ _ - _ - _ _ - _ __

l l

I i

APPENDIX A Listing of Corrections to be made to the SAR  !

4 l Eggg Section' Change 1

p 14-16 4.10 This change will. clarify.

l: the difference in the type of drive used for the j standard and transient l

[ rods. j 4-16,17 4.10.2 The paragraph is modified I to reflect the new step- j ping motors used in the j control rod drives.  ;

1 l

4-16b 1 Figure 4-8 The figure has been up- ,

dated to depict the new  !

control rod drives.on the standard control rods.

4-22 Section 4.11 The phrase "three ion chambers" has been changed j to "two ion chambers and a pulse detector" to allow a Cherenkov detector or an ion chamber to be used for pulse operations.

22 Section 4.11 A paragraph describing the l NM-1000 has been added to the SAR.

4-22 Section 4.11.1 The section describing the Multirange Linear Channel has been updated to re-flect changes incurred by the new console.

4-23 Section 4.11.2 The section describing the Wide-Range Log Channel has i been updated to reflect 1 changes incurred by the ]

new console.

]

i 4-24' Section 4.11.3 Portions of the section ]

describing High Flux  !

Safety Channels One and Two have been modified to reflect changes incurred by the new console.

_ _ _ . _ _ . _ __.._.__________.._.._____.___.____m__ _ _ . _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _

l.

k Pagt Section Change

29 Section-4.11.4 Portions of the section

describing Fuel Teapera-ture Safety Channels have been modified to reflect o changes' incurred by the new console.

, 4 Section 4.12 The RWP associated with l

the wide-range log channel I in any mode other than I OPERATE is no longer i

required.

See 10 CFR 50.59 writeup.

4-27 Section 4.12 The SCRAM associated with any of the safety channels in any position other than OPERATE in no longer required.

See 10 CFR 50.59 writeup.

e

.{'

i l.

APPENDIX B_

Soecific-EAB.yazd channes for the neviousir discussed Facility Modification Safety . Analyses i

i i

l 1

.I

1. REACTOR CONTROL COMPONENTS (Section 4.10)

CURRENT SAR WORDING:

" Control rod movement within the core is accomplished using re.ck and pinion electromechanical drive for the transient control rod."

PROPOSED SAR WORDING:

" Control rod movement within the core is accomplished using rack and pinion electromechanical drives for the I standard control rods, and pneumatic-electromechanical drive for the transient control rod."

i

2. STANDARD CONTROL ROD DRIVES (Section 4.10.2)

a. CURRENT SAR FIGURE:

Figure 4-8 PROPOSED SAR FIGURE:

Figure 4-8 (modified to reflect new control rod drives)

b. CURRENT SAR WORDING:

"The standard drive consists of a two-phase motor, a '

magnetic coupler, a rack and pinion gear system, and a potentiometer used to provide an indication of rod I position, which is displayed on the reactor console."

PROPOSED SAR WORDING:

I "The standard drive consists of a stepping motor, a magnetic coupler, a rack and pinion gear system, and a potentiometer used to provide an indication of rod I position, which is displayed on the reactor console CRT."

c. CURRENT SAH WORDING:

" Clockwise rotation of the motor shaft raises the draw tube assembly."

I PROPOSED SAR WORDING:

"When the stepping motor is energized, the pinion gear shaft rotates, thus raising the magnet draw tube."

I I

I I

j

':?] .'T .'.a: .~ : T;?: ,,

7 I

gg gge g 081 YEN 0704 I adlCROSWITCMES SPRf886 LOADED PULL R00 POTENTIOMETER S

k.- ,

I y e DRAW TUBE

b ARMATutt 7,PO,P0 00 00, ELECTA 0mAamE7 BASSEL Pil798 I C0808ECTiesG 300 g a "

p ALWh8488Und CLAO I ,

pr 80AAT80 0AAPweft COstTROL A00 sulfM EBLIO ALutet8sume 80LLouv8A I  !

I .

i g .

g ,,0 ORE o STANDARO CONTROL ROO ORIVE FOR SAFETY AND SHIM ROOS

• I

% Reut Dame asumer Esemat Orsue tesear Cassemens Temsumuseer 4-16b i

i

)

I

.  : ;;E: 3 A.:.  : - 'I '; l I PoTENTICtLETER LEMS gg , DRIVE MOTOR l 8 MICR05 WITCH E$ l SPRINGLOADED PULL R00 I .

cr 1P $

. 1 .

I j

~

DRAW J ( )

et ARMATURE i

j TUst

.l s

! {

l TIP OP PUSN R00 FOOT f

I ELECTROMAGNET BARREL Pf570N I 4 l I 9 CONNECTlHG A00 l

I

= h i i

l

[ ALUMINUM CLAD BORATED GRAPHITE l '

CONTROL R00 WITH SOLID l ALUMIMUM FOLLC'WE.9

, OR AIR FOLLCWER

[

I .

I FIGURE 4-8 Standard Control Rod Drives I

I I 3. REACTOR INSTRUMENTATION (Section 4.11)

I I

1 l

CURRENT SAR WORDING: {

"A fission detector and three ion chambers comprise the )

remaining detectors."

PROPOSED SAR WORDING:

"A fission detector, two ion chambers, and a pulse I detector comprise the remaining detectors." I l

4. NM-1000 ADD TO THE SAR: (at Section 4.11)

"The NM-1000 system, which includes the Multirange Linear Channel and the Wide-Range Log Channel, is contained in two National Electrical Manufactures Association (NEMA) enclosures, one for the amplifier and one for the proces-sor assemblies.- The amplifier assembly contains modular I plug-in subassemblies for pulse preamplifier electronics, bandpass filter and RMS electronics, signal conditioning circuits, low voltage power supplies, detector high-vol-1

{

tage power supply, and digital diagnostics and communi-I cation electronics. The processor assembly is made up of modular plug-in subassemblies for communication elec-tronics (between amplifier and processor), the micro-processor, a control / display module, low-voltage power supplies, isolated 4 to 20 mA outputs, and isolated alarm outputs. Communication between the amplifier and pro-cessor assemblies is via two twisted-shieldad-pair cables."

5. MULTIRANGE LINEAR CHANNEL (Section 4.11.1)

CURRENT SAR WORDING:

"The multirange linear channel reports reactor power from source level ( 8 thermal watts) to full steady state power (1 MWt). The output of the fission detector, fed through a preamplifier, serves as the channel input. The I multirange linear channel consists of two circuits:

count rate circuit, and the campbelling circuit. For the power levels less than 1 kilowatt (t), as selected on the power range select switch, the count rate circuit is utilized. The count rate circuit generates an output voltage proportional to the number of pulses or counts received from the fission detector. Hence, the output is proportional to the neutron population and the reactor power level. For steady state power levels at or above 1 l

11 l

1

l l

l 1

L kilowatt (t), as selected on the power range select switch, the campbelling circuit is utilized. The campbelling circuit generates an output voltage proportional to the l l reactor power level by a verified technique of noise j envelope amplitude detection and measurement known as )

campbelling. The output-from the appropriate circuit is ]

l fed to an amplifier which supplies a signal to the strip )

l chart recorder located on the reactor console. The power level is scaled on the strip chart recorder between 0 and 100 percent of the power indicated by the power range select switch on the console. The strip chart records this' output for all steady state modes of operation but )

not during pulse operation.  !

PROPOSED SAR WORDING:

"The multirange linear channel reports reactor power from source level ( 8 thermal watts) to full steady state 1 power (1 MWt). The output of the fission detector, fed i through a preamplifier, serves sa the channel input. The f multirange linear channel consists of two circuits: the i count rate circuit, and the campbelling circuit. For.

{

power levels less than 1 kilowatt (t), the count rate <

circuit is utilized. The count rate circuit generates an l output voltage proportional to the number of pulses or counts received from the fission detector. Hence, the i output is proportional to the neutron population and the reactor power level. For steady state power levels at or above 1 kilowatt (t), the campbelling circuit is utilized.

The campbelling circuit generates an output voltage proportional to the reactor power level by a verified technique of noise envelope amplitude detection and measurement known as campbelling. The NM-1000's micro-n ocessor converts the signal from these circuits into 10

.near power ranges. The multirange linear channel output is displayed in two formats. These are a bargraph indicator on the Reactor Control CRT display and a strip chart recorder located on the left-hand vertical panel on the control console. The power level as displayed on the  !

CRT bargraph and the strip chart recorder is scaled i between 0 and 100 percent for each of the 10 linear power l ranges. The multirange function is auto-ranged via the NM-1000 control system computer. The multirange linear output on the CRT bargraph is displayed for all steady state modes of operation, but not during pulse operation.

1 i

- _ - - _ _ _ _ _ . b

(.

j' l

6. WIDE-RANGE LOG CHANNEL (Section 4.11.2) l CURRENT SAR WORDING:

lE .E "The outputs of these two circuits are log amplified and then summed in a summing amplifier. The summing amplifier i a supplies a signal to the strip chart recorder located on

.g the reactor console. The power level is indicated on a 10 decade log scale (10-8 watts (t) to 1 MW(t)). The strip chart records this output for all steady state modes of operation but not during pulse operation.

During certain steady state modes, the wide-range log channel also measures the rate of change of the power I level, which is displayed on the period / log meter located on the reactor console."

PROPOSED SAR WORDING "The outputs of these two circuits are digitally combined and processed to provide the power rate of change (period) and the power level indicated on a 10 decade log scale (10-8 watts (t) to 1 MW(t)). The wide-range log and period outputs are both displayed on bargraph indicators on the Reactor Control CRT and on hardwired vertical LED bar-graphs on the left-hand side of the Rea"+1- Control Con-sole. The outputs on the CRT bargraphs .re displayed for all steady state modes of operation but not during pulse operation."

7. HIGH FLUX SAFETY CHANNELS ONE AND TWO (Section 4.11.3)

a. CURRENT SAR WORDING:

"High flux safety channels one and two report the reacter power level as measured by three ion chambers placed above I the core in the neutron field."

PROPOSED S AR WORDING:

"High flux safety channels one and two report the reactor power level as measured by two ion chambers and a pulse detector placed above the core."

I b. CURRENT SAR WORDING:

"The steady state power level, as measured by the two high I flux safety channels, is displayed on two separate meters located on the reactor console."

1

l i

PROPOSED-SAR WORDING:

"The steady state power level, as measured by the two high ]

flux safety channels, is displayed on two separate bar-  !

graphs located on the reactor console."

c. CURRENT SAR WORDING:

"During pulse operation, high flux safety channel one is shunted and the sensor for high flux safety channel two is  !

switched to a third, independent ion chamber placed above the core."

PROPOSED SAR WORDING: I "During pulse operation, high flux safety channel one is shunted and the sensor for high flux safety channel two is ,

switched to a third, independent pulse detector placed '

above the core."

l l

d. CURRENT SAR WORDING:

"The NV channel output is displayed on the strip chart recorder located on the reactor console. The NVT channel /,

output is displayed on the reactor console NVT meter."

i PROPOSED SAR WORDING:

"The NV and NVT channel outputs are displayed on two separate bargraph indicators located on the left-hand side of the console."

e. CURRENT SAR WORDING:

" Knobs for each channel, located on the reactor console,  !

allow the channels to be checked for calibration.

Switching these knobs to any mode from operate (i.e., to the zero or calibrate positions) causes an immediate reactor scram."

PROPOSED SAR WORDING:

" Calibration of each safety channel is done automatically when the operator initiates the " pre-checks" by activation of the Prestart Check Switch on the control console's Mode Control Panel. Any failures detected during the prechecks >

will be automatically reported to the operator via the reactor status CRT. This calibration can only be per-formed while the reactor is in the SCRAMMED mode."  ;

1 1

I

-m i

h l

.I L f.

h CURRENT S6R WORDING: . .

1

y. "A trip test knob.for each safety channel ..."

I V

! N PROPOSED E&R WORDING: J "A trip test switch for each safety-channel ...

l

[

=

8 '.1 FUEL' TEMPERATURE SAFETY CHANNELS-(Section 4.'11.4)

L a.. CURRENT SAR WORDING:

"The two fuel temperature signals are amplified and displayed on two separate meters located on the reactor console.. During pulse operation, the output.of fuel I temperature safety channel one is also recorded on the reactor console strip chart recorder."

PROPOSED SAR WORDING: j "The two fuel temperature signals are amplified and

-displayed on two separate bargraphs indicators located on the reactor console and on the reactor control CRT."

b.- CURRENT EAR WORDING:

"A-trip test knob for each fuel temperature safety channel, located on the reactor console, provides a means of testing the scram capability of each channel without having to actually reach or exceed the technical specifications limit on allowable fuel temperatures."

PROPOSED SAR WORDING:

" Calibration of the Fuel Temperature Channels is done automatically when the reactor operator initiates the

" pre-checks" by activation of the Prestart Check Switch on the control. console's Mode Control Panel. Any failures detected during the prechecks will be automatically ,

reported to the operator via the reactor status CRT."

9. RQQ WITHDRAWAL PREVENT (RWP) INTERLOCKS (Section 4.12)

CURRENT RAR WORDING:

"RWP prevents any control rod withdrawal if the wide range log channel is in any mode (i.e. position) other than OPERATE."

PROPOSED R&R WORDING:

-This requirement is deleted (See document for analysis).

)

l l

)

l l

l- 10. SCRAM LOGIC CIRCUITRY (Section 4.14)

CURRENT'SAR WORDING:

"Any of the safety channels (fuel temperature safety channels and high flux-safety channels) in any position )

other than OPERATE (i.e., CALIBRATE or ZERO) causes a j reactor scram."

PROPOSED SAR WORDING:

-This requirement is deleted (See document for analysis).

)

i

)

l I

I I

I I l

I i

1 I l l

t 1

I-I E

APPENDIX Q I AFRRI TRIGA Console (Safety) Scram System Single Failure Criteria Analysis I

I I

I I

I I \

AFRRI TRIGA Console (Safety) Scram System Single Failure Criteria Analysis l

REFERENCES:

1. IEEE 279-1971 Criteria for Protection Systems for Nuclear Power Generating Stations.

2. um 379-1977 Application of the Single-Failure Criteria to Nuclear Power Generating Station Class IE Systems.

The following analysis is postulated upon the principle (explained in Reference 2, Section 6.1(4)] that redundancy of protection devices provides complete assurance of safety in operation with Tegard to the parameter monitored by.

the device. For ' example, the failure of a fuse to blow when subjected to its designed rating of overload current is a credible possibility, but the failure of two identical fuses in series to blow simultaneously is not a credible possibility.

1. The steady steady-state timer scrams the reactor after an elapsed time and no rW=cy is provided. he probability of the failure of this device is estimated as follows:

Mean Thne Between Failure (MIBF) of the electronic circuitry is about 200,000 hours based upon parts count and stress factor per j MIL-HDHK-217B. At 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per month this is one failure in 83 l l years.-

The electronic timing circuits operate relay contacts whose failure rate is expressed in operation cycles rather than MIBF. A conservative estimate based on manufacturers specifications is 25,000 operating cycles. At two cycles per day and 5 days per week, this is one failure in 48 years. The most likely failure is increased contact resistance rather than welded contacts so that an unsafe condition probably is not credible in less than 100 years of operation. The steady state timer is not a required safety  !

system component. )

I I

2. The pulse timer scrams the reactor after completion of a power pu!se ar.d I no reAnad-ncy is provided. ne rated life of this device is 250,000 l

electrical operations which exceeds the probable number of pulses to be produced.

The probability of random failure calculated as MTBF per MIL-HDBK-217B based upon parts count and stress factor is greater than 300,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. At 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br /> per month, this is equivalent to one failure l in 125 years.

3. The manual scram button is used to shut down the reactor manually. The specified life is 100,000 cycles of operation. At 15 manual scrams per day this would be one failure in 25.6 years. However, this is a normally closed switch with a direct acting operator. Le most likely failure mode is a broken switch structure which would result in failure to reset after a scram. Welded contacts would be separated by mechanical force of the direct action operator. R AaaA="cy for a manual scram exists in the console operator key switch and power on switch. '

4. The console key switch de-energises the magnet supply as well as other circuitry. The estimated life is 10,000 operations. At 15 operations per day, this is a failure rate of one every 2.6 years. However, the key switch is not depended upon to perform a safety function except to prevent unauthorized startup. The ==aa=1 scram button provides shutdown redanA*acy so that an unsafe failure is not credible.

5. All reactor tank shielding door interlock switches and emergency stop buttons remain from the existing system and are unaffected by the new hardware. The amargency stop switch and all other switches on the new console use the same actuator and switchmg element as are used on the existing system. "

2-i

6. The loss of AC power causes the magnet supply to be de-energized which in turn produces the same response as a msnual scram, dropped rods.

7. De high level trips in the two power safety channels are redundant and therefore do not present a credible mode for failure. All non-safety outputs are physically separated and isolated to prevent comnon mode failures which may otherwise invalidate the single failure criterion. A minhwn separation of six ' inches, or a metallic flame barrier exists between all safety and non-safety circuits. A minhwn isolation voltage of 1500 volts RMS or DC applies to both optical and transformer isolation.

The MTBF of the two NP1000 safety modules is greater than 20,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> based upon component failure rate data taken from MIL-EDBK-217B. The bistable trip portion of the NP1000 has an MIBF greater than 200,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. Because the NP1000's operate independently, '

each with its own detector from the existing system complete reA=A="cy I exists.

8. The detector high voltage is interlocked by trip circuits in the power and safety channels and the reda_a A*at circuitry makes unsafe failures not credible. Separation and isolation criteria of item 6 above apply.

9. The two fuel temperature safety channels are high reliability modular signal conditioner / limit alarm devices each with calculated MIBF figures exceeding 200,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. He channels are r-A=A-at with separation criteria applied to the wire harness therefore an unsafe failure is not credible.

10. De magnet supply ground fault detector uses a high reliability modular signal conditioner / limit alarm. The signal conditioner module has an MTBF of greater than 200,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />. He limit alarm uses a relay rated for more than 25,000 operations. Dere is a pushbutton switch which is used to test the operability of the ground fault detector on a daily I .

basis. l Because the relay only operates during testing and tauit conditions the end of life cannot be reached. Therefore the probability of an undetected ground fault is the probability of random failure in the signal conditioner which is less than one in 23 years.

11. Pool Level Monitor - Pool water level is monitored with redundant float operated switches and redundant relays with contacts in the scram circuits.

The switches and relays have failure rates of less than one in 106hours but redundancy makes a water level monitor failure not a credible failure mode.

12. Watchdog Scrams - A watchdog timer on the data acquisition computer and another on the control system computer are required to be reset periodically by a program routine as a safeguard against computer emnponent failures either in hardware or software. If the required response is not received within a definite time period, rad'=A=rt normally open (fail safe) contacts interrupt the scram loop dropping the rods and shutting down the reactor.

The watchdog timer is an additional safety device.

1 1

4-

_ _ _ _ _ _ _ _ - _ _ - - _ - - ~

1.

l l

I APPENDIX D_

Scram Circuit Safety Analysis for thg University gi Texas TRIGA Reactor I

I I

l

u

.t? .

> ,@N COLLEGE OF ENGNEERING H" -

1. p' e $

e C! "A THE UNIVERSITY OF TEXAS AT AUSTIN DepatmentofMecharicalEngineering NxdearEnginorringProgram Austin, Texas 78712 (312)47131f6 I

. April 22, 1988 l

(-

1 Mr. Junaid Razvi l General Atomics P.O. Box 85608, Ms/21 San Diego, CA 92138

Dear Junaid:

As per our discussion at the TRIGA meeting, I have enclosed a copy of the complete safety circuit evaluation we developed from the available GA information. . I hope that this analysis might provide valuable support for your analysis of the new console installation. A review by knowledgeable persons should be made to ascertain that our understanding and evaluation of the documents is correct. I believe that although the system has evolved from some of the documentation we had available, the changes to the analysis are not likely to be significant. An effort was made in the method of presentation to demonstrate various conditions.

Please review and return comments. Other persons have also expressed an interest in the analysis but I'd prefer to have General Atomics comments to make available on final document.

Thank you for your help in this matter. ,

i Sincerely, 2.

Thomas L. Bauer Assistant Director Nuclear Engineering 1 Teaching Laboratory TLB:div Enclosure i

'I EIJe Etnibersity of Ecxas at Gustin Scram Circuit Safety analysis for The Unluersity of TeHas TRIGH Reactor Prepared by:

Dr. Thomas Bauer Professor of Mechanical Engineering David Goff Engineering Science Student l

April 22,1988 b

I

TR IG A -IC S l Reactor Safety System i F rot +;tiv+ h 'fle ni af th+ E+3;ter I0f+ry :yit+m IR;;) ar+ r r vi't+d tal ifV ritl par 3rnet r 01+SIUr ni+nt :11311H+1i Bn.! O C'. Ittr?l-rul p/a../+r :1f ~ Ult j I (3-lTalu Circuit) El':h ni s!Ur+111+11t ell 31111+1(0LitrOli Operit10!i 0I th i .'ritn1 CitTult by fn+Sni Of a r lay in tus i:1f.'Ulf V/h 11 iny 011+ Of th is l'+13ys !!

l j

(

tripp+d it cut 3 p?t 7+r to th+ 0:ntr01 r?di Ih :<rDrn (trCUlt d il{n li (Ornpfl5+d ?f fc ur iLinctiollal ! :ti: rii Th+e r [r i nt Th [r?t'<t1V SJ.(10I1 Ut? nit::ritij Of th+ iyit m, UlOlilt0!i U1 Int lyit+fn i Op r81:111ty, i 10f13 /ST+c 911d n1111091 S+:t1011, !!1:ludi!1% the };+y f 3 ilt<ll and n13nU31 i<r?!1li,311d th ph*/liC31  ; ':!rcuit iti+1f,1ncluding th+ prOUrld (Sul';

and p01,cl+r iupply rnc.nitors. Thes+ i+:tions are ihuwn in th+ disgrant b+10w.

Frate: tin Actx.n ENoblJ (* ' LGCP 6,y ;,jict, i r Fr:;: sm i +

~~

• f Catrel mg F+1ris 20VtC i t.

I ivurce cvcon itstus

) : R+I N '

! 5 9?

Control

": Greur.J $

p ,. ,3 l L Protective Action signalJ (-) L0op RSS Functional Diagram I Th+ following analysis first lect 3 at th+ basici cf th+ syst+m in g 5tvady-state op+ ration. Aft +r a g+neral failure mod +1 is d+v+1cp+d, th+

analysis +::pand; to look at th+ cahbration checks, th+ bypass r+1ay us+d in puls+ mode, and monitor chann+1 failur+1 cutside th+ scram circuit itself.

I

l i s r

l L l u . 1 RSS Failure Analysis i;

Th+ RSS scram arcuit inppli+$ pow +r to tn+ control rods an.i h+nc+ ts th+

point at whMn all i< rams <=: cur, er fail to eMur. Its prop +r function 15 (11+r+for+ imp +r3tive tu iSfY op+r3tton Of th+ r+3G)r In analy:tng r3m rh+ ic

.ar:utt. ni many pot +nualisilur+ mod +s as pos!!bl+ w+r+ +::armne:i to

+iamac+ che pr-:.babtitty :t a ar:un isaur+ The ulumat+ fattur+ 0:n!+qu+n:+

UEi that th+ ?ntr?1 rc :15 + /+rt nor in!+rt+d 3nd no !cr m Occurr+d luring 3

!:rsm iltuation in order t: +::Smin+ th+ my m t/hich Individual n1191 -5 m tr.+ arcuit might 1+1d to a non-Er3m, a (Sult tr++ tas construct +d tM+.'i on an analyiis of th+ i: ram arcuit.

• Th+ fir!t st+p in th+ RSS failur+ analysts myolv+d id+ntdying th+ various ways in which the f.iS could fail. Th+i+ includ+:

1) Physical Syst+m Fatlur+

2) Limiting Safety Syst+m S+tting (LSSS) Failur+

3) System Op+rabl+ Failure

-4) Comput+r/ Manual Control Fattur+

Th+ Physical Syst+m failur+s includ+ wir+ br+aks, shorta, and fSilur+ of tn+ iround fault d+t+ct and voltag+ d+t+:t arculta Th+ LSS3 (3ttur+$ ar+

n'ti+ which wculd c3us+ loss 0f th+ at flity to d+t+ct an unsaf+ c?ncituan.

Th+!+ +1+m+nts includ+ th+ Fu+1 T+mp+ratur+ monitors an'. i:'+ F+rc+nt Power monitors in th+ N!..!-1000, !!P-1000 and NFP- 100v '; it+m Op+rabl+

failur+s ar+ tho5+ which caus+ loss of th+ ability to monitor th+ op+rabl+

condition of oth+r syst+ms, for instanc+ th+ high voltag+ monitors. Fmally, Comput+r/ Manual Control failur+s are tb. !+ asso.:iat+d with th+ program r+1ays or the manual scram and t+y switch.

Th+ fallur+ analysis is ba!+d on a fault tr++ approa.:h in which th+

probability of a particular failur+ is brch+n down into comp 0n+nt parts which

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - _ _ - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ~ - - - - - - - - - - - '

f-1 I

r I 3r tth r 3dd+d 0T nlultipli .:1 CO{ t!i+r d p+ndtng 0n t/h+th r (ti+

l C0rnp00 nt.i IUnCC).:'n in An 'Or" Or in Snd" Ininti r r ip+<t1V-ly Th+ g+n+ral .

+quation for th+ fault tr++ II Pleil=f t = P Af.g g : PL.,.

- e f .ag

+

• P . ,...k ' f t p,

,w.) ' (1)

Y'/h -r F ri,i.re li' IU* 0YOI3ll IIODSDIllIY 0 III'"ilI'? Ult f 3111Dj TJ FrMD in 3 17.rM11 iltuhtl:n 3rr.i th f : 3r th+ pr0L,SLility Of +i:h Of th+ fillUr rn':d i I

g om ,r d sm.:..

I I

I -

I I -

I I FAULT TRCC OVERVIcw I

C ur.er A F,.2 F + l+ n;+

F silor, I c utrat Circuit input I n, s se ., . -

g .I nta nt Fallue e F hi1ur+

I k I L555 Enr.t $9st+ri' Cor.iput+r /

C +tectiora Opr st.1+ Fr.9 sicin F nik.r* Oprator fint+rn F alur+ p3 % p g, )

I =P i

P +P -PComp /Han + PPhy$gs Failure (1)

I Sysop L$ss I

l I

l-1 l-L l

l Physical' System l

Th+r+ 3r+ m9ny p0t+ntt31 f 3119r+s in (n+ phy1101 syst+m F0rtun3r+1y, rre:it r+ivit in 1. :i :( F0t /+r to rh+ '. ?nr.r01 r0ds and h+nc+, a icr3m :ttV3nOn

/

Th+ poi!!tl+ f allur+ med+i are Ih': f t M lin '!"[pl" (0 t +(Urn!  !

f'?t<!+r !< ii Ih0f t C0 p01..? r ! 20V ['C ind + (0 + Or - t0 -)

Ch61t (0 lin tiu[ ply to IUpply Or r (Urn to r turn)

!h:'rt to 2r00nd Or0und d+ttct circuit fa!!UT+

Chort to p0W r b to - Or n6t 2Ql/ ['()

Fow+r fluctuation I/Olt!'g d+t+ct circuit f ailUr Th+ first tw0 failur+ typ+5 inh +r+ntly icram the syst+m by cuttmg cif p0w+r to th+ control rods. Ther+fet+, th+y ar+ n0t Of conc +rn for this 3naly!!s. A short along +ith+r th+ supply or r+ turn train or to a pw+r supply which !! simliar to that suppli+d m th+ scram circuit would not b+

d+t+:t+d by the scr3m circuit. Su<h a sh0rt would n+ght+ th+ Saf+ty r+13ys b+!0r; th+ short if it w+re in th+ supply train or thos+ aft +r a short in the r+ turn train. How+v+r, for this to 1+ad tc an ur. .afe failur+, such sncrts I w0nld hav.: to occur cn both th+ supply and r+ turn trains b+cavi+ all 13f+ty m<nttors ar+ duplicat+d on both trains. This r+dundancy structur+ is shown i

in the fault tr++ and mak+s this a non-single failur+ m0d+. 1 Th+ Oth+r branch of the fault tr++ shows the probability +s ass 0< tat +d with f Sults in th+ ground d+t+ct and voltage d+t+ct cir vits. For th+s+ to caui+ a pot +ntial n0n-scr3m situation, h0w+v+r, a short to ground muit 0ccur as w+11 as th+ ground d+t+ct failur+. Similarly for the voltag+ d+t+.:t circuit, only i s+nsor monitor failur+ coupl+d with irr+ gular voltag+ can caus+ 3 p0t+ntial non-scram situati0n.

l

l l 1

i l

l l

Ih+ ':[Q3tiOD I?T tlili $+j!h DC OI th Iault tr ,th rt,ii.

I g#t,5;ss .e or F4 ar . eir 6ete:t . e v F4 .de . gv &ctut . #;h Lw.e . e;t r...rIa c.,

t/'lh r+ th6 Equir6d t6Trn 1Gdt:0t i that 41th+r a ih0rt tc p0w+r Or .Sl?ng i

th4 lint rouit Occur Gn L0th th+ iupply and return lines li E. P.,,,,,,,1:, th+ i pr0bability Of a ih0rt t0 p0W+f t.'ht.h 15 .hif+r+ht frGrn th6 p0W+r supply On.1 h te;+ dst +crabl.- l tt th+ V0lrag+ rnOnit0 ring 01rcult5 Whil+ P;Lt.eer tith +

[r0bibility Of d ih0rt to pot..'+r ui:liitingulih3bl+ froth th+ p0W r iUp[ly I P,e: n 30 b subititut d int 0 E.1Ulti0n 135 I/ Art Of th OV rall failurv pr01: ability.

I I

I I

I e

I I _ _ _ _ _ _ - _ . _ . _ _ _ . _ _ _ _ _ - _ _ . _ _ _

PI-h'5ICAL SYSTEM FAULT TRec i

F t..p14 21 I p t +r.*,

I till  !

l l

n <

I;Il

,1 i:

. I q vr. .. ;:.1 i.,,.,,,,,

F Sult , 1

' '" t r I  !

i ..

I'f)I Q Fo ve +r G, cur.J Shcrt Shert I ler.itcr licr.itor on Out cr it.

F silur + F *ilur* L it.+ Lir.+

l l 's a

\

fb [h l l Po v +r Faw+r Orcur.J iir. ort ihcrt short ihos t it.<,r t L+tu tor L,ppl., L+t.ctor to ts to to ,o F ntlf f ailf F ntis Gesear 1 l in+ p. . .. +c Lin+ pe, ... +c Phytys Gr.f ault

• Per, o,, + Pp ,,,,,

• P p ,,, + (sh P . tin.+P,,,,,,.P (21  ;

1 i

)

l i;

I l

)

l~  ;

Limiting Safety System Setting I.

Th+ LSSS consists of th+ fu+1 t+mp+rstur+ mcnttori and tn+ p+r.:+nt pot!+r montturi F?r +1th+r httu fu+1 t+mp+rStur+ Or perc+nt poi.t+r to

('aut+ 1 n9n-t<r3rd !!tV3t!On relD'/1 On br>th (11+ f upiply and l'+tu rn (rging  ;

mV *f IS11. This is b+caui+ ther+ are two in.i+r +n;i+nt f u+1 teny+ratur+

rnO!11f 0ri On+ (9nne.t+d to e3(11 line <.'.I th+ !.:ran) (tr<:utt, ?gtnllarl*r (11+ r +

hre ; p r':+nt p?' ? r ni.'n!?nri Ind+[ ndently .:6tiO+ 'I d (O (!;+ 3.:r)n) ;1r :Utt

tlat m ,5rd+r icr L f Silur+ to

Ur bc th '. ic.uld h:m+ t:. fail. Tlai ti .'l+1rly 0 F n-!!!181+ I!51111re G1/>d .

Ih quit!On I?r f.h prob 31: 111!7 OI l.".I311Vr ?ti 'iliOWC In th I3Vlt tr+ 15:

I *0- u;; a (*L'r t. ,)2 , ip in.r'i =:

' 2

( J ')

g Fu,, may be plugg+d into Equation 1 as part cf th+ ov+rall failur+

probabihty equation.

I e

es I -

I I

I - _ - _ _ _ _ - _ _ _ - - - - - - _ - .-

1

]

i 0 1 i LSRS FAULT TRCC a.,

i

{8 j ,

i L

4 Lili i.;i t at:,

F uIz I

j, Ib

.i 1

i

+

r en1 i it; RFa nu

, Ci+ t+ction Ce,t+c ticn F ails Fuli 1

.L

?

A fu+) Fu+1 EP se 81 EPwr* 82 T+ rop. Tunp. Fuls F uls a i r r,tir a: Fiiir 2 2 P

LS5$

=(Pfiemp 1.gp R Pust 3 131 l

1 i

\,

,,__mu__a--- - " - - - - - - '~ ^ ^ ' ' "~""'--" - - - - - - '

1 4

1 1

l System Operable Failure Th+ syst+m op+ribl+ compon+nts ar+ th+ 10'v/-5.4/3ter 1+v61, hqh watchdcg, and +::t+rnal scram r+11ys Ea:h of th+i+ has ind+p+nd+nt i+niers wir+d inte both tn+ iupply and r+ turn'!!n+3 and 3015 a non-!!ng!+ failur+

rned+. Th+ 10V-w3t+r 1+v+1 m0nitors v3t-r 1+v+1 in th+ tank. High v0 l :h+:1:s th+ "Oltai+ On th+ p+r:+nt p0* r+r m.:, nit 0ri Sn:

1 th+ +:2+rnit : rim i

ini .:r+5 th9t all +:7-rnal . .:.n 11tt.: r.i ar+ rn+t. II 3pph :St.1+ Th+r- Br- npair i

}

t '. rat.:h.10i' r+1Byi :n+ for th+ C _;: Ond On+ f:r th+ DAC Th+y rr.'.r.!t?r th+

1.:g r.'13r- .

)

n.I ' /t!! . . ram if n0t r+i+t +'l+ry f!v+ i+ :0ndi by th+1r :niput+r Th+ +quatt0n 70"+rntng th+ prcbabihty aisectat+d with th+ syst+m l Op+rabl+ s+gm+nt of th+ fault tr++ ts:

1 I P.we - (Puv )2 + (Pm)2 + (Pcasa)2 ce + (Ps 12 + (P ,e12

/. )

~

)

Vih+r+ the iguar+d terms are due to the redundancy in th6 syitem. P 1 ses  !

Can D+ plugg d i!sto Equaticn 1 as part of the overall failure probability.

3 i

I I

I '

,I I l I 1 I

1 i

i L

I o cu ra 2ISiL v C.r,a < rt 7.n0_t r nU ar , c3 . __ __

_ .xtc

).-

l-  !

I t

5) !ITfil 0p +f' bt l+

F s nlar+

Ih l

I l 3 ;. r .. v. c,, ,, a. .

E. +c i., t u 3,g , m, Trit. F hilt t..,4g j F ,iis l'i) l 1:.')

ll l

E t ,c:,,1 r,,,,,,.,

C'*,C heni" ('e f +< ter ('+ t +C t.*.f*

,t.2 >; .i %.;n . _ F hIls F ,1\s F illi FiilJ

(- [.,

(n*)

Q_ _

c .,c = l Exte m ) .t+ m l I

'csc a'llcic =2 04"2

'I W " I "Cf W "

Fuls ll Emily Fiils F ail; Le ., V ater Hign F4ils f ails Lev el Trip Volt n p F siis Te,c. p n1, f,

Ub

_Je id o P<. ,9 sop 2

= P( se + Pone + Exs P' PLWL2p HU2 III to, go, g ,y , , , g ,, , ; ;

V n t +,- Vit+r F ails 7,iu l

=1 F nili " F nilt l l

j 1

I 1

a i

Computer /f,f anual Control This s+<ttan d+scrib+s th+ p Obability of tailure Of tit + progr3m r+1Byi D. net an Optritor ICr3m. It!V+ th+ pr08 rant r11375 are id+nt1<al, th+ p0$51bl+

]

p fa!!ur+i ar+ tn&t on+ relay fatti to op+n on c0mmand, or that two, thr++ or all f.:ur f 3tl II only On+ relhy f3113. Ini rtt0D of th thr+ r+rnhining rlJ!i titil shut dOLin th+ r+3:t.:r 10 tut 315 n0t an uniaf+ fattur+ rnod+ If sny U % thr++

Or s!! f0ur r+1ay1 fit! t0 Op+n, th+ r+3: tor will net shut down. It 11 +3t!!y d+m0nitrat+d with a pr0bab111ty tr++ analy5ti that th+ pr0bab111ty of f 311Ur+

Of ~;,3, or 4 cf the r+1ays is 6P f 2+sP3+Py4 t wh+r+ Pr is th+ probabutty of a Iinil+ relay failur+. This +::pr+ss10n will cl+3rly b+ dommat+d by the first t+rm f0r small P 50 t th+ cub + and fourth pow +r t+rms will be disr+ gard +d in furth+r analysis.

Th+ 0p+rator scram is normally initiat+d with th+ manual scram switch In

~

the ca!+ of a switch failure, how+v+r, the op+rator has oth+r m+ans of I shutting down the reacts.,r. Th+ss ine.lude the 1:+y switch and th+ individual rod contrcls. The 4::pr+13 ion for r0d contr01 failur+ is ta5+d on th+ iam+

l thr++-Out-of-four logic as th+ program r+ lays as again, only thr++ rod 5 mu5t b+ in!+rt+d t0 Shut th+ rGactor down.

l Th+ +:@r+ sion, th+n, for th+ probability of failur+ of th+s+ subsyst+m; is.

E.:. /m. - 6P,,,% 2 , (p,,, . P 4 6P,,,,,, 2 ) (c,) )

Ilot+ that th+ op+rator has thr++ ind6p+nd+nt m+ulcds to scram th+

l syst+m, all of which must fail for a non-scram situation to arts +. This is highly until.+1y as the switches th+ms+1v+s are r+dundant. Th+ manual scram switch, for +nmple, is wir+d dir+ctly into th+ rod control circuit at two plac+s.

i l Both 0f which muit f111 for th+ manual scrim t0 fall. Cimilarly, th+ 1:+y l

swit:h it wir+d dir+ctly into the scram circuit and also will 3+nd a pow +r off

iignal to. th+ . _ _ GC. _ . Titti

, . .signal

- it0p; th+ GC fr'.>tn Updating th+ W3tch.-10g t,

' tirA f? AIF.i alt T. flyi.I,f 5,..,II.'li, th y y/11) (ini Out !<ritnnittig tut s'tr<utt if th+

s

itre':t r+1Dy f at!++;t..co.dq fo Fin.1117, th+r+ 3r+ th+ in.:ltvidual r0.! u. l.tr.:!s l- Th+i+ ar+ run through tn+ G Enci so d-Insnd tnat th+ tc ftt.nr+ t.+ 0p+rsting /

- !e..r0p.+rly; hot. ratr+r, th+ t;rgt :h;lOg r+13yi 3r+ .lailgn+d t.: 3, r un e q+ : 11 HR In .

I th V nt Of i !? f t' 'ir iltl0T 4.iiuniing th n th3C th+ 50f tt /s .i runntni, e

.Only Chr++ ^i Ch f00f D:ilientf0li rni.Ji( II.'!iCC10n [r0p fly t0 ihUr cl?t./n th+- ,

r b<t0r,1+ch+r? 3@ 3in th+f+. nil)it IM tW0 f 3ilUr i for th9 Syf t+1n n0t to icratn. j l

Av+rall, th+n tn+r+ rnuit t+ i+v+ral catastrophic failur+5 all cuurring 1 q

illnultan+0uity, non9 of WhlCh 13 Chui+d b7 an V tit Which WOUl:1 tr!{g+r Oth r $3f+ty Syst ins, for th+ Optrator n0t (O b Sbl+ to Scrani thy 175t+1n. l Cl+1rly, th +'.'pr $$ ion tS donlinat&d by thy ChinC6 Cf a prograin r+13y i

fdklU[h dnd tIlh h[0b$b11II 0f tIIY chh[htDI I'hknf Undb!h k0 6C[S[6 th+ $ $( ID iS VaniihinglV, $1n311. '

. i 1

)

1 l

1 l

.t i

(-

l l

l i

I.

g g COMPUTER / MANUAL-g FAU4_T. TREC. .

C a. . + l rtin I i aliwre i

1

!\

i I

g 3,g I

.4 it . < >c +

$cr ir.)

ik1is

A' i

L  ;.

I ". i 1;

E,bcLups 9 "#,' ','

1. " 1 R+149 2,5 er 4 F Fh ls to I f.t'J Switch Rg,d

~

vG+6 R,lny; r til l

Cor.trels C,o+s Not F 6113 Not n i F nil Pre v +rit Etr.gle 1

St.ut Cro w r, F nilur+

d I , ,;,,,

, , ,,, t , a, . , ,,.. .

s I Op+n i

C,oes Hot.

F sil i

Nc t n Pr+v+nt g ing),

$r.ut C> owr.

F ailure I Pg

,,,f g,,= 6 Per. nelav + (Pg3,3,, *P g,,,y 6 Paodcfra )

I I _ _ _ _ _ _ _ - - - _ _ - - - - - - - - - - - - - - - - -

L L: Failure Analysis 1

= =r; 4 .lvlat.iy.0f the.r+1ays-in eft + Erain.<;usult ar+ of th+ sarn+ typ+ and ll+nce H:. 3,_hav+ tdenti:al f 3tlur+ probabrittles -Th+ lugh 70ltag+, p+rc+nt pow +r, tow

- Wy+r, Wa.t<hdogdu.+1 C+mp+.ritur+ +n+rnal scram and pr0 grim r+13yi i'rt j all finular. An +::pr+i110n for th+ +itunat+ l failur+ rat + (0r r+1ays ti f00n.l in i..httra r'i H3 n 1b. 1: _17 E+um. n E It is I:-)i .! on In+ +nvir:.nm+nt, .*y:1+i p+r h00r th3t th r 157 li +:.) s:t ;.1 (0 0F Tbt I'n:! Of 0:Uri , r lay ty[ .

Th Handt+:t.11v+s th+ +::pr-filon for inlur+ as

'A' r

• b ( bl.

• E4 'Ec'Ecyc Ei' Eq)I3IIDIYI/Ib bfi$b)

?.iiUrning a d0uble pol +,-liingl+ thr0w,501 nold T+1Sy 6p+ rating St 155 thSn 00+ cycl p+r hour, carrying less than fiV Amps, th+ 11t+ratur+ giv+$ the modification factcrs 35:

p, = -t.6 Environm+ntal Fact.:,r

.p,=1.5 Contact Type Factor p,y, = 1 Cycl + Rat + Facter pf = 12 . Family Construction / Application Factor pq = 15 Quality Rating Factor pg, = 1.2 8 : Load Factor j

& = 406 ' Bas + R+1ay Failur+ Rat + 1 i

Equation 6 th6n giv+s Ar = 1 failur+/106 hrs. If Pgis the probat.thty ci a

])

1 relay failure per hour, th+n Pg = 1 x 10-0failur+s/hr. l

)

For th+ rnanual ?.:rera, control t0d and 1:6y switches, a similar +::pr+3310n appli+s:

A 3 = b (p, ' p, ' p,y,

• pt) f ailur+1/10 h hrs (7)

Where:

p, = 2 9 Environm+ntal Factor

p. 2.0 Contact Typ+ Factor

+i L

I:

l l p,.,, = 1.0 : Cycl + Rate Factor pi= i 77 : Load Factor 5:. ..c. ~ = ig sWi!F9PisFSwffch'Fritilt+ P.at+ " ~~ " 4 *

. :.:. .: .= . -- % . -

2.Tfi+n A3= ..=::L C

=:, . . . . . . .

e. failur+5/

...e.

IU

' hti and3P = .. :: 16" failur+s/hr. Ikt+ that this is cnly th+ probability of 3 f.hystfal !)tlur+ cf th+Rittch its+1f. Housv+r, b+c avi+ ci th+ r+dund9ncy in th+ op+ rat!<..n of th+ switches, as deicrtbed in th+ ?+ction on op+r3 tor scrarni this probabt!!ty !! much larger than that of th+. 5 wit h op+riting prop +rly, but Ialling to scr3m th+ syit+m du+ to int +rnal iyit+m (a11U1.

For th+ conductors in th+ ctreutt, data is given by the IEEE Guide to the Cell +ction and Presentation of El+ctrical El+ctronic $+nstn7 Comrenent and H+ h3nical EquiDrn+nt Reliability Data for Nucl+ar Pow +r G+ner? tine Stations Th+ Guid6 sui 2+sts from empirical data that for a short to ground. th+

probability is Pg =' 1 x 10-7 failures / hour /10 circuit f++t. Th+ probability of a short to pcw+r is Ppwr = 6 x 10 4 ailur+s/ f hour /10 circuit f44t. Itis asium+d that a short to lina is similar in probability te a short to pvw+r. '

The ground and voltage d+t+ct circuits w+r+ assum+d to have the sama l failur+ rat + as a s+nsing instrument over all. This is a rather conservative +

numb +r th+n, as th* d+ tact circuits ar6 much simpl+r than most sensing instrum+nts and have f+w+r failure mod +s. R+ liability and Risk Ansivsis sugg+5ts a failur6 rata for a s+nsing instrum+nt as: Pinst

• I % 10' failures / hour.

1 Th+ protabiliti+s calculat+d in th+ !ault tr++ analysis th+n, giv+: -

Pg,7,o, = 5

• P a2 = 5: 10-12 Pg ,g;3 = 2 ' Pa = 2 x 10W Pe m/na = 6P g2 + 4P34 = 6 x 10-12

s EPr. vin

• P a# Piu t

• P; ve ' Pa:t

• U E pve N

• 2 % 10~ I '

_.. ;.. _ . . ._ ._Using ti1+1+ nurabers In Equation 1. Lt+ 3++ th:.*. . _ .

= - " -?7 ggr.ly_1 l_tmore-A Qu;cr_12i+3n tin 1+ b+t'./++n failUr+; ..f 1::107

)

i

- - = y+Ers For thFfallUr+s con 21d+r+J: Itis trnportant to not+ that tini is not th+ i

"a

t+d (11n ICT (11+ Ctr' Ult (0 20 t /1th0Ut failur+, rli+ I?ng lif (Ird+ 15 r3th-r e

indic)t.tVY Of th+ Inh +r+nt deil4_111f the syitsul In th3t 31151n2[+ failur+i '.../111 ,

' Euf+ 3 5013!n '0!i.lif.10n, thtr+f01,0111y U.VO >r n10ri f al!Urts occurring ilmUltan+0U?ly 02n 193d (0 3 p0t ntially Un531+ f311ur . Th+ unprobability of tuli h3pp+ning i$ r+f19Ct9d in f.hY lON I111UT pr0b3b!!1ty I

I I

I I -

l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ J

i

)

L l l- i j

Bypass Relay l L

j

. .: . Th+ t,ypass c.. : .

r+1ay is ui+d to cut th+ HF-1000. cut of th+ scram circuit upon w.:q. 3 +. .nt+ ping puls+' '-hiQT+:.".Whin, this . occurs, culyicne.. monitor for p..erc+ . . . .

, r+rnairis abl+ to scrim th+ iyit+m. The prec++: ling analysts on fattur+ mod +5 q

th0W3 that on+ of the r+3 sons for th+ +xtr+ n+ saf+ty of th+ syst+m is tn+

{

r+dundant'/ inh +r+nt in 011 monitoring sy!C+ms. This r+dundancy 13 l l

<0naprOmis d wh n the r+xt?r gc+s into puls+ rn0d+ Fortunat+1't, the r+MtOr n0rmally 5tay5 in pul! mod + f0r 3 V ry sh0ft tim so th chanc 0f 3 failur+ at that instant 15 V+ry small.

A pot +ntial pr0bl+m could ari3+, however,if th+ bypass r+1ay i >l fai1+d and the syst+m did not r+ turn from puls+ mode. In that + vent, tne . I em  !

could op+ rate for an extend +d p+riod without th+ NP-1000 to provice .he ,

l extra safety factor. If the bypass r+1ay does fall, how+ver, this failure will i

b+ apparent on th+ operator's display. The percent power indicator for the HP-1000 will r+ main blant b+cause the CSC will not be receiving any I information from it. It is, ther+ fore,important that th+ operator ch+ct the HP-1000 display each tim + the reactor is pulsed to insure that the bypass l i

r+1ay has return +d the system to steady-state operation. l i

Not+ that +v+n if the bypass r+1ay fails, the NPP-1000 is still monitcring the sy3t+m and would be able to scram the system should the perc+nt pow +r

+xc++d its limits. For th+ circuit to remain in operation and totally unmonitored, the HPP-1000 would also have to fail. This again cr+at+s a situation in which two failures must occur for an unsafe situation to arise.

Th+ new probability +quation for the 1.555 du+ to the bypass r+1ay is:

l Puss = (P, ,,,,,,)2 + (P,,,)2 . (p r ,,, p, ) , 3(p, 2) , 3g i o-t 2 Inst +ad of Pg ;, = 2(Pa ) = 2::10-12 as b+ fore. i This still giv+s an overall Pnu m =lx10-1I f ailur+s/hr, or a m+an time b+tw++n failures of 1:107y+1rs. '

i 1

1 I

(

i Appendir: Explanation of Equations  !

l Th+ +.quations giv+n for swit.:n and r+1ay f a!!ur+ ar+ of 5tmtlar form.

1..

\

- .. ........,..---u--.. , .

.. _Th+y mcluda.a. .tas+

failur+ rat + for th+ giv+n component typ+ 0,) and 5+v+ral modifications (pps) ba!+d on "= :ylitiidual,:ompon+nt and th+

1 iyit+m in which it 0p+ rat +5 Th+ modtitcation factors us+d ar+ +: plain +d k b+10w -

p, Environm+ntal Factor  !

p, C0ntact Typ+ Factor i p,y,e Cycl + P.at+ Factgr p, Family Construction /Appli:ation Factor -

pq Quality P.atmg Factor pg, Load Factor i

lium+rical valu+s for th+ p(5 ar+ given in Military Handbo2 217-P.+v. E and have t+en transcrib+d in part. Most of the modification factors depend ,

cn whethsr the componsnt m&+ts MilSp+c standards or is considered ' lower ,

i quality' In the interest of 1:eeping failure +stimat+s consGrvativ+, it is j assumed that compon+nts ar+ not MilSp+c quality.  !

P, is basc-d on the environm+nt and installation typ+ For a fi +d ground installation, p,is 2.9 for switch +s and 4.6 for r+1ays.

Pcis the same for relays and switches and depends on the form and number of contacts. Values for Pc are shown in Table 1.

Table 1,. Table 2 Table 3 I!!DA 1 4 Ratino f.c

.05 Q

5PST 1. 0 1.02 A .I DP1T 1.5 .1 1.06 P .3 iPoi 1.75 .2 1.28 M 1.0 3PST 2.0 .3 1.76 L 1. 0 1 4P$i 2.5 .4 2.72 Not Rated 1.5 l DPDT 3.0 .5 4.77 3PDT 4.25 .6 9.49 4 POI 5.5 .7 21.4 6 POT 4. 0

_ _ _ _ _ _ _ _ _ ~ _ - _ - _ _ _ _ _ _ _ -

For pt, , th+ load f actor, u31o+5 3r+ .1+t+rnun+d by f, wtuch is tn+ ratto of I. . . the 14ad.curgent.tt the r.ated resistive load. P ,t value6 for an inductanc+ based

.......-..:...a.2---.. -

.c. . . . .: . - *~ ~'.- -

I J{F.ZMl+nold Mlay3ratfoynfrin.TSble-9 abov+. The r+1ays ar+ assuni+d to t.+

.- .-- 7_~r~nt+'d leir* 120V whldh s.if+i an S = .2.

For a swit:h, p,.,,,, is +qu'ai t0 th+ nurnt+r of cy.:1+i per hour that th+

switi:h ti ep+ rat +d (p,.y: = 1 if 1+io than ley,:1+/hr) F .r r+1ayi. p,n,, is 1.0 if th+

r+137 ?p+ rat +5 St 1+!i than 10 :y:1+i p+r hour.

It+ quahty f 3:t:r, pq. it ihown in Tatl+ 5 Th- r+1ay rningi 3r+

ur.1:nown and h+nc+ ar+ assuni+d tc t + ur. rat +d.

Finally, p,.,, 15 ihewn for i+veral ri-lay con 5trucacn typ+5 in Titl+ -:

t.+10w.

Table 4 l ,P_t Contact C urrent Construction 1921 I 8 3

18

$lgnal current Lois munit and mumps Armature Dry Reed Hg tuotted I

8 Hagnetic Latch 1 -l Soleneid 0-5 Amps I

6 Armature 10 Balanced Armature 12 Solenoid B

= Th+se factor 5 can t+ plugg+d into Equations 6 and 7 in th+ failur+

Snalysis to g+t:

A r = 4 ( PL P,

• Pc

• Peye

• P " Pq) fatlur+s/100 hrs (6)

A.r . 006 (l.28

• 4.6

• 15

• 1.0

• 12 ' 15) l Ar = 1 Failur+/106hr3 A.3 .g (p,

• p,

• p,y,

• pt,) fsilur+s/100 nrs (7)

).3 034(2.9* 2.0

• 1.0

• 1.i3) l A. .:< . 2,

- Failur+s/100 hrs

Calibration Checks 2

m:.n W At:systhissir.41p-4he411tration~oi=s+veral'syst+ms is ~ch+<h+d

= ~ - - =- ::.._..._

~

Vdd, '. ailtomatic311y,. Ih+it sy4tsnit trerhigh voltage monttor5

p+rc+nt pow +r

~ monit?d,T6+1-t+mp+75tilisthonit&5,'a'ndlh+ Ohkhdog tim +rs Th+ 10t?

.watYr 1-Vel, +::I+rnal scram i+ttings. manU31 yr3m 3wgrch 3nd l'ay tytt t yrs not t+ic+d by th+ auto pr+t+it and thould t+ ch+ct+d manually.

Ih j' E nt pow r IU+1 C 111p ritur+. and high V0lt4gY rnonttori Sr+

.:n+:t+d by m+3ns of r+13y1 wn!.:n iwitch fr?m th+1r normal p^ittions to cut th+ monitori out of th+ syst+m and allow a t+it current to b+ run through the trip 5+ction of th+ syst+m. Th+ CCC monitors wh+n the syst+m trip 5 to insur+

that it is at the sp+<tfi+d point. Th+ r+13ys th+n r+ turn the syst+m to normal op+ rating mode. To check th+ watchdog timers, the CSC s+ts +ach tim +r and mak+s sur+ that it times out at the appropriate time.

For th+ high voltag+, perc+nt pow +r, and fu+1 temperature syst+ms, if any r+1By fails to r+ turn to normal op+ rating mod +, no curr+nt from th+ d+t+: tors l would reach the monitor circuits and this would result m a scram. If, how+ver, an entir+ system + g. the fuel t+mp+rature monitors, fails to r+ turn to normal mode and the calibration current remained on, the monitors would

'I not scrarn but the d+t4ctors thems+1?+s would b+ compl+t+17 cut out of th+

~

syst+m. This is obviously an und+strabl+ situstion. Note that th+ only s.-ray for luch a failur+ to occur is for the CSC to 1+3ve th+ calibration signal active a i

1 and fail to r+ turn th+ calibration relays to th+ir normal op+ rating posttions.

Mer+1y leaving th+ r+1ays in the wrong positions will cause a scram when th+

calibration curr+nt is turn +d off. l If both of th+ss failur+5 occur in ons of th+ high voltage /p+rc+nt pow +r l monitors, th+ calibration voltag+ will b+ pr+sent and show up Bs variations in p+rc+nt pow +r and high voltag+ on th+ op+rator's di; play on th+ CSC l

e

l: .

l (assummg that the calibration curr+nt dG+s not +::.:++d the syst+m limits and Taus+ a.,sg[a[Otsili.} A15A, since.th+ calibration of '+dch monitor unit i+ J

~" ~ ~

t.Mii!@fQt)Npend+'rttl5.tNti.must fail IN tl$(sy.s.'t+m as awhol+ to op+ rat + in an unmonitor+d mod +. If th+ failur+s oc.:ur on th+ fu+1 t+mp+r3tur+

l m0nttors, th+ CC diiplay ihould ag3in show variations du+ to th+ <altbr3tton

.;urr+nt. How+v+r. th+i+ uniti Br+ ch+<t+d all at on.:+ 50 if th+ 5ystem I3 tis, I th+r+ 15 no bad.up sy5t+m and th+ fu+1 t+mp+ratur+ r+ mains unmonitor+d. If th+ V:ltage < :ntinu+s to ramp as it do+s clurmg th+ cahbration <:h+ct, though, it should quickly trigg+r a s.: ram on its own.

Ther+ ar+ basically two failur+ mod +s associat+d with th+ wate:hdog tim +rs: failur+ to r+54t and failur+ to tim + out. Both of th+se rneles ar+ t+ited in th+ pr+-start calibration chects by simply s+tting the timer and 1+tting it tim + out. Ev+n if th+ CSC g+ts stuck in th+ calibration mc l+ it is a safe f allur+

as in this mod + th+ CSC waits for a time out after s+tting th+ tim +r. W+r+ th+

syst+m in op+ ration, th+ first such time out would caus+ a scram. Th+

waundog tim +rs could also be reset by a random tignal. but this is unlik+1y is two pairs of tim +rs would require a reset. There ar+, then, no unsaf+

fSilur+s asso< tat +d with th+ watchdog timers' calibration.

Th+ additional failure probabilities for +ach subsyst+m du+ to calibration of th+ sy3t+m ar+ assum+d to b+ those of th+ each subsyst+m failing all at onc+. Th+r+for+, th+r+ are two t+rms to b+ add +d to the ov+rall failure

+quation, one for th+ fu+1 t+mp+rature and one for the p+rcent pow +r/high voltag+ monitors. Th+ t+mp+rature syst+m has thr++ relays which must fail simultan+0usly and sach NP unit has two r+1ays which must fail simultan+ously. .

P.g g ,= Padu *Pm = Pg *2 p 2g= Pg =4 1::10-24 Failur+s/hr

. i I

P r .T. = Pg 3

= 1::1018 Failur+s/ltr SP-' __3@[17 t4tW.thMMYJRirCraM}[ssjfr),+rs tf,r.(ig,nk[hE 1 mil 1+r than ;' ' *

.65._.__.

W 1+. _Mor..t1I+'.syit+m.!G fhc3 a AMtF. Tlif}htildt sigtufiIanby aff+ct th+ 'v+ rad -

[ ~ . f ailur+ probability. _,

i d

i 1

)

4 e

'l

. j

i I

I l

{

il Monitor Channels

---in atiition to th+ scram circuit. wtts+1f, saf+ty_jyit+m failur+i could occur _

. a ....

.-- .- m a. : m --

I ;~u ... ... .

. . ,r='; ?7in tlif 61Frutors th+mGlM ThFmonitor chah'n+1h of s.p+ctficanport arr t i

~ '

i 95T$t tinip+'riture nib 8itori and the HP-T00dhiEl flFP-1606)+r:+nt pow +r /

[

1 --

hiih t.*:.itai+ monitors os th+I+ ar+ critical to th+ saf+ Op+ ration of th+ syst+m.

t' Fc r this anal' tits. th+ <hann+15 ar+ all assum+d to hav+ th+ instrument failure rat + ih0wn in th+ abov+ analy;15 and all failur+s are assumed to t+ uns3f+

l This ti a c n5+rvativ+ + stim 3t+ as som+ 'mmon failur+ mod +s, + g loss ci itgnal from th+ d+t+ctor, t!ould caus+ a s : ram.

The instrum+nt failur+ rat + is giv+n by Pinst = 1 : 10-0 fa!!ur+s/ hour.

g Nc,te that this failur+ rate is th+ sam + as th+ failur+ rat + us+d for th+ relays in th+ circuit its+1f. F0r an unsafe fuel t4mp+ratur+ failure to occur, the analysis 15 id+ntical to that for th+ scram loop itself i.+. both must fail for the system to b+ unsafe. This leads to several permutations of failures which are unsafe.

.I However, all require at 1+ast two failures. The original expression was Prtmp

= 1::10-12 How either the monitor or the relay can fail, but one must fail on

+Sch chann+1. Therefore:

Primp" (En + P )2 , 4 z t o-12 f ailur+s/hr.

g Cimilarly, for the NP1 1000 and NFP-1000, the added failure modes increase the number of possibl+ failur+s, but th+ syst4m redundancy still protects the system. For the NPP-1000,in addition to the monitor failure, a f

gain failur+ is considered. The NFP operates in a s6parate gain moda for I pulse op+ ration and w+r+ it to switch to pulsa mode during steady state f 1

g op+ ration the NPP would essentially b+ useless as the trip point in pulse 1 mode is much higher than for steady state. Since the percent power and high voltag+ failure ratis are incorporated into different parts of th+ overall f atlur+ mod 61 and the p+rcent pow +r failur+ rat +s are also aff+ct+d by the bypass relay, it is +asiest to loch h+re simply at the incr+ase in f allur+s -

I.

i.

e L

i

<3us+d by consid+ ring th+ rnontr0r Chann+1 f 411Ur+i. A detailed analysis is

.,. -- pr+i.+nted

- in th+.f0110 Wing.+::a.rnpl+.

_ . - - -- .-__ . . -- - Th+-additional failur+ probabtitty.

% ,7~ - - ~ ' ..e.; .vz'. TZ.a -

W

• O 4" 5 * * .== ..

• h_ - ~ ' ~ ^"t l:; 2 :. . r.-ten;4derint=.tistnt-metic.ntf=~ the byp,a,ss4r+1ay *and NPP gain turns out to t.e:

._~,..'~' '

9 :a c,2 Pmmma -- .-- - -=.-6z=-?  : 10-1 7 +2 failur+s/hr.

!~~ .

. Thisis sss46tially an increas+ cf 1.l':: 10-11 fattur$s/hr and tring th+

o't+rall failur+ rat +, inc0rp0 rating th+ bypais r+1ay and instrum+nt Iallur+3, tU

_ 10 !! fattur+i/hr. Thii 9';+1 a rn+an tim + b+tw++n f ailur+1 of *,:: 100 y+ars, ll:t+ that thii nurni r li li ntially d0ubl that I0r th bali 0 Syst In, Wh!.:n 11 to I:- *:j' :~ d ai th instrufn nt Chann li <0niider d had Sirnilar failur rat i IU th+ r lays in th+ UTCult it3 lI.

t l

1 f

w-_

• ge$,

,mm--

Analysis Example Th+ follovfing.is an +;2ng.gt+-inalytis11shtin tlinifalI0r 3; n - c

~_

7~ ' . ..-.-

-u..--= - m ~ ;-

_ , _ . ~ . - -

r__ -

, . ,- ~[D;;;;irlwkinfat tTiFidrGfit pow +f syst+m, th+r+ are six fa

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

can .

- .. .1.__'..c.aui+

IjPP-1000 monitor, the IIP-1000 perc+nt power scram r+1a hrt.uniaf+-51tuation Th+s+ ar+

~

^

y, the IIPP. t 000 p+r:+nt pow +r 5: ram r+1sy, th+ HPP-1000 gim mod + r+

rnod+ bypais r+1ay In all cai+! I311urs .:fr two.:ompon+nt -

o caui+ an Unmontror+d ittuatt0n but ne t 111 failur+ uch pairs will r+

j a attuation. Cinc+ th+ NP and HPP ar+ on dtif+r+nt im+s, on+

must fail in each i.e. an NP monitor and NP scram r+1ay fattur+ 1 combination as th+ liPP-1000 is still fully functional. The table b+1 illustrates the possibl+ failur+ combinations.

hPP,,M h_PP-g hPP,=,[

P ,Nf-M NPP- M - Mf. -} Susass S

$ U U U NPP- A $ -

S U i U U NPP- 8 $ $ -

U U NP - M U U U U -

NP - A U U U $ -

Bypass 3 U U U $ $ -

NPP-M: NPP- 1044 Moniter NPP-R: hPP-1000 $ cram Relay NP- M: hP- 1000 Monitor NP- R:

NPP-4: NPP- 1040 Saia NP- 1000 $ cram Retag 5: Safe f ailure i.e. system still monitored typass: typass Relag U: Wasafe f ailure, system not monitored Th+ tabl+ clearly sh0ws th+ Incr+ase in failur+5 from th+ o~rigi ,

which had a p+rc+nt pow +r failur+ rat + of 1 1012(NP-R and HPP-R in the tabl+). There ar+ nin+ unique Iallure mod +s shown rease of above for th+ inc 8 x 1012 discuss +d in the monitor channel section.

~

Conclusion

__.e .. disstat%I2+/pr.e this_afaly.lis.giy+s an ov+rall. failure probability of 2 x._.

p .

e...

J.,!-M,_pJsifur+e peg..Inu.irdhis givss.an appronm'at+"rhun time t,+tw++n.

@/M I.=.dailur+s.0f,.i.::..,10'/..y+ir5dhspit+ th+ s++ ming +xtremity of this dumb +r, it ..

d-r r was:Stt+mpt+<1 throughout.th+ analysis tb m31f+"all assumpti0ns as .

i:0ns+rvativ+ as r+150nably poliibl+. Th+ inh +r+nt r+dundancy of tn+

lyit+m simply m9tes it highly improbabl+ that ans failure would destroy V

if th+ mt+grity of th+ iaf+tyyjitAnt! b WWe c 6Q c(np L -

2.[fL*l i

tar u.afn n ry L ....ukus a C .he .4 .

( yKtTdi~s ~ oint, a cothparison ol'thiTifety systFm's,5113rit!'ity t6that 61 p s th+ physical syst+m its+1f might b+ cf int +r+st. Ptltabillt't and P.isP. Anelvsis giv+s the failure rate of an individual control rod physically sticking as 1 x 10-4 per day,1 +. i x 104 failurss per hour 'ThM r.=.t.i 5 m,tua; ) ;;";;.g th. / ;.'.r;' ad th L, .nt m u. ?, um, .u 6 25 un.. 0.. : . . :';r :y W

. hm ., e de7. %'+r; th. . ;;.cta anu ned ;- op +4 oW v;uy u3 i u .m a

':1,* 't+ ":Ury' f21. . ew ovud be tL vs tuuve u.shwi. Using the thr++

out cf four logic that only thr++ control rods must function in ord+r to cause a scram, the probability of failure +quation is idsntical to that shown for the program r+1ays in the Computer / Manual section and is dominat+d by the '

t+rm c*P/ . This gives a failure rate for just the control rods as 1 x 1010 f ailur+s p+r hour.

Granted that this number still provides a r+assuringly long m+an time b+tw++n failur+s (1 x 106y+ars), the point is that this small s+ction of the physical plant alon+ has a failur+ rate which is almost an entir+ ord+r of magnitud+ gr+at+r than the failur+ rate for the +ntire R+ actor Safety S'ist+m. Clearly, the Reactor Saf+ty Syst+m is one of the more r+1iat!+ parts of the r+ actor d+ sign and is not likely to b+ r+sponsibl+ for any syst+m fallur+sde seaws.

)

)

i L

l.

L 1

APPENDIY g Analysis 9.f. Elyt Dollar Raan Inmartion QX1r. & IER Second Interval l a 1.h.1 AERRI. TRIGA Reactor ~

k

Revised 4/26/88 ANALYSIS OF 5 DOLLAR RAMP INSIGtTION l

l OVER 2 SECOND INIERVAL IN AFERI TRIGA REACTOR I,

J

{

l 4

I l

I Work Perforsned for ARMED FORCES BADIOBIOLOGICAL RE!NEARCH N ,

Bethesda, Maryland by GEMDtAL A2VMICS under Contract DNA004-86 C-0011 Amendm at P00005 l

April 14,1988

l l

i AFRRI RAMP ACCIDENT l

l Summary - With the computer controlled TRIGA Mark F reactor the control rods can be operated in a bank which makes it possible to add large l

amounts of reactivity in ene action. The speed at which the rods can be withdrawn is a variable parameter. An accident scenario is postulated I

such that during a startup, the following sequence of events occurs:

1. The transient rod is fully withdrawn preparatory to going to a steady state power; 2.

The shim, safety and regulating rods are t' ten withdrawn to establish 1

criticality;

(

l l

3. This withdrawal occurs at a speed which would withdraw the total rod-bank in two seconds from a sub-critics.1 condition; and

4. The safety systems terminate the excursion by scr===4ng the reactor at 110% power, i.e., 1.1 MW.

The consequences of this accident are trivial. The maximum fuel 1 temperature is about 3300C.

Although the excursion results in a peak power of 340 MW, the reactor power is below 1 MW in less than 1 see after the initiating event, i.e., the beginning of the rod withdrawal. In Fig.

1 there are shown the results of this accident.

Analysis -

Use was made of the computer program BLOOST3, a lumped parameter neutron kinetics, thermal-hydraulic program. This program has bean used extensively in the analyses of reactor transients in which reactivity changes are rapid and the event is of short duration.

In Table 1 there are listed the reactor parameters used in the analysis.

1 I

i TABLE 1 Reactor Parameters, f- Initial conditions:

No. of Fuel Elements 87 Core / Coolant Temperature 25'C l

Initial Power 0.01 watts cold, clean excess 3.5% 64/4 ($5.00) )

i Rod Wortha j Transient 2.56% 6s/K ($3.66)  !

Shim 1.30 ( 1.85)

Safety 1.30 ( 1.86)

Regulating 1.27 ( 1.82)

Prompt neutron lifetime 39 psee Fuel element specific heat (C+7T)

C 821.7 joule /*C 7 1.67 joule /('C)8 Core water specific heat (per element)

C, 860 joule /*C' Delayed Neutron Data I $ 1 (see.Q 1 2.310 x 10-4 1.244 x 10-4 2 1.528 x 10-s 3.051 x 10-s 3 1.372 x 10-s 1.114 x 10-1 4 2.765 x 10-s 3,o13 x 1o-1 5 8.049 x 10-4 1.1362 x los 6 2.940 x 10-4 3.0135 a los The integesi fuel temperature coefficient is shown in Fig. 2. The coefficiw a itself is approximately 1 x 10-4 AA/A*C. The coolant temperature coefficient was assumed to be sero since it is relatively small and, also, because in the excursion little heat is transferred to the water.

With only the transient rod withdrawn the reactor is suberitical by 0.37%

6s/4 ($0.53). The withdrawal of approximately 10% of the rod bank occurs before criticality is achieved (based on a normalized s-curve for worth

vs length withdrawn) so the 3.5% ds/s (SS.00) insertion occurs in 1.8 sees instead of 2 secs. In Fig. 3 the reactivity inserted as a function of time from the point at which s = 1.0 is shown.

{

Since the transient is terminated when the reactor power is 1.1 MW (110%

full power) only a portion of the 3.5% ds/s is inserted at the time of the scram. A problem was run to determine how far the rod bank was l withdrawn when the scram occurred. The reactivity inserted in the ramp l

was 1.305% ds/s ($1.86). This represents about 34% of the rod length.

l To this must be added the 10% withdrawn before criticality was achieved.

Thus 44% of the rod bank length is out of the core and now participates in the scram. This portioe of the length represents 40% of the worth of the bank, or 1.55% ds/s ($2.21). The total scram activity is, then 1.55%

+ 2.56% ds/s, or 4.1 1 ds/s ($5.87) total with the pulse rod worth added I to the banked rods. The rods fall under the influence of gravity in L see from full out to full in, following a delay time of .015 secs to6 allow the magnette field to decay. Since the rods are also influenced by the resistance implied by the passage through the water, the rate of insertion is not as the second power of time. If there was no resistance the rods would fall from full out to full in in less than 0.3 sec. By assuming a resistance term that is proportional to velocity and that the drop time from full out is 1 see, the reactivity inserted as a function of time frois first motion is shown in Fig. 4.

Conclusions. The postulated accident scenario in which a bank of rods worth 3.87% of ds/s is withdrawn from the AFREI TRIGA Mark F in 2 sees, with the safety system functioning, will cause no damage to the reactor or harm to any person .

l l

_ . ~ _ . . ..

~

. . . . . - - . _-. . . _ . . *6L.

. .t d.A ad. ~. . .,r.rras

. ,. .. , . . t. . ,

e I

= = . . . _ . . _ . . .... ..... . .. ...

(

1

.. .. s... .

.._...s...

g

. .., . _ . _.n ......

g e , .

)..

. -.6..-._---

4 , i

.3 j

l ,. . . . . . s .  ; .

...9 . .s 8 4.6 e ,

e

.=. - d._.. _

t ,

..e..

i .~ . _ _ . .

4 ...,,.,>.2._,t.. ..

. - . . . . g s . . g ,

.:. .g .. v - . .. ..

_e... s.. .

._=_,

s i

~

c ,

t . 1 i i

! . .,. .t .

y.

. , g- j 4 i ,

........,.....,..,.._<ge..,._,,,

....t. ..

. a

- - --- p ,

a.

**'.s.._.. .

c....,i._.. . w. .

i.' '. ..._-'..} .:. i.:...

!...t. .I j.. . .j;

, _ , _.,qiT.T

. .L . .. ' - .i .: .. ...- .

r _._._. .

.. .. ..i.

. . ...s.. . . . .,., . . .s, .. . . ;

W . . .

, . _ ........g. # g W ..=.4 .p._ ...

. ..- __ !.. .......a..

....4...

. . . . .l.. .. .

. . =...a_i..,,'.l .. . .. .. 6 . .l . ,,

...w.. . ..... ,

~ .. . --- .. ."I...,, ._..

I to -

y - , , , s ,.

, l es e yr .

... .. r  : .! in g

• uQ_,

s. i *  ;..  ! 4 ' L. .~ 4 - 'i ' -

9. w - V .- J. * - li. ;s ,

, e t s. n s..

. t . , . . . .v

.3  ;*

y 9 1 i [- *

• 1 8 I. 9 346 [ /. .** . , 8 .r 3 -;

) 9 ,. ,b.: { . .i- . 'l *

,t i l ^.L*.*f. 76.*1 i j . g' '

ce Q/ _ . v.

. .. .. _. J '.7'

_ _ i .. .j _ f.

ft.,1 * .g.....a....>.6.'.'.,. ._t

..7 .

. ' .t?. " V r

• 4 j . . -

g 4. _ _ ..

{ , q~ - =~

+. 9. .h.ewle . -.-.. ...

. , _ . , r ' .J i ,

n: t ,

. -l 3 p , j M. .

M.

..g

.e. , #

, j .,,

t 6 . = . *. . .. , ... .

e. ... . ..

. t ... .

_.. - _.. . .. == _ . . . Ku. g,.% , . . . -

Df. .. . ... .. .; g. -.

t . . . ..L  ;. .,

3 _ . . . *fta l% .,

f, , G.4

,_ , , =in. % w#

N. . . . i. . . . . . . . .

1p6 %.

. -.v% s

._ ~

.u, g

.. _ _ - , . 1 L. - .,

.s 1 9

s s j ... .

Y,'N i .

, j, i , * . t. J

..,d

. 1. . _ _ _

v g i

i :

-k 4 .

. i -' ' ' ,

.s U .

. _ . . . . . .. ..,..j .,... .. .'......3.. .>. *

.w....!.

s

_ 6 _ _ . . .- _ _ _ ,

,' . 4 .

i

. .4. .f 1 l s

'kr 8 '

, 1'swV. DOM p.) 0,61, .

/. . - .._

- .. f ._ .. *_ _ :_10

.%. . .. g, _....._......_........_._s..

.+ ..

.. ,.\

(

4

_. ./. . . .. . , ..%.f. 4M. ._ . . . _ . . . . ._ _

S.

\ 4 4 s

/

l

.s. .m dD $. . . _- - - - _. . . . .,

.._._f..g,.=_g )

9 l- 4

. . . . ... . .p .. [

s

.- _ .M/.S ._ .

. . . . . _ _ .._ . . _ . . . . . - ...i . ..'._-..__..3..- i

!, \

.. .. g -. . . . . . . . -.

. . . . . , i

.. . ._ ../... _ _ _ . . . _ _ . . _ _ . . .._.__ ... _ _ ..,._

\

t

~

I ,

pp n aae

\ '

o.t p

'fo, 96-Tm mn hw Geo k*W M

_-.___ _ - _. dh CJkh% YY

l  ! '

i _ i l l i l i de 1I b \ mi(M.f4 k g,.d I l l :f?.l I' l l 'l  ! l ! l 3 l l l l-: 1.

-lL  !: -l.. j y l "[ .. j [ j ]

1 l l i l j l l l l

- . " h" I r!: d::{N*. .' E 4*:b _~} l ,', 7 ' gi l'I  ! . " !- - l l 'l -! .}' i l l -!

a i lW ._:= Q .l=si_ of . :: ..-\,( r ,. T I . !; 4-!: i; i. ' .I _ . l .. - l  ;

-F. ;r s = t=B==isWEEs==EJ N=lk :-0:=!=-l'.-!#1=J;T4=i! Hei-ii Jl uI J i i

_. .m . _ := === ==r. _._ _ _ _ .=w -. _.; t .-* ._n -l* - =. gr r--?--- -- t .:..--.

' . .y;MJ . . _.. . : .r .:rt,.:: ;u:..l 1= . u . . .

v - .i

=-t = =--t W :c. . . h _:.= _=r - *-- l -~ ": t :. - :f __ ._; -

-t- - : t- --J

- " .._N  :- .:_..h.

- -- - ~ - - --- -- --

+ ,--: up

= = =_ _-q=== _. i'.>..i . . .h .=_v-_ f:===,

.=v=_ _ . - a

. :* . - .=t..I*:.

.=:4.J.-  :  : m-

.a -

r-- ~ ~=

-- --1' ____: -l:__._.- ,_ =_ __-

. .=v.= s =- : = Y= == :1==:_==ut-.m =. : . ::=::. :-=. .- r =.=.l.-. _._. . . .. :,: :=.: . - -._ . I j -

=. r=~ .: s --: - -.-- - , ~~,_---=_l=._=_.=_=:,:_.===r__=:---:

-+ = r :-  : .= = t= = :t

=. ._

_4 _= .

- : .- 9

% - _- r.- - - -

. t=._

___ i _ c =. . =:: . _ __ ; =- :_-:==

-  :._ :_ .- = _. _: ; . . . .

,1 4 0 ___. -

. _ _ . - _ _ _---. .___ ._.__.___=_.__.....

i

_:,w

. __ .=t- -=.__=_=___a_.._-_. ._ . _

y ~

_/

~

.e, n  ;-=.=___

_L:=.:--

_ 4,=:=__.ca ._-_ -- - -;_ ._=..up: =::: :q.: . =_- . . . . /.-

N :_== --

". ~

~%__:-- d. =i .5='/ *:5* =

h $.v_ ..=_;=- - -

. ._ _ ~ -;--

_. _ . _~_. _=;= .=.=-

==:  : - .-

. =; =-= =-. :- -f .r.- ;_ _.a. _ ;

O, --

~> .

.=_._.au_.__. _ . __

___-. _ ;.=_ c_._ r -- . ~+

___=4,.=..==_=4; n -. f :=l ..

J =__ .:

, ,_,-i = -

-_r.-_._..-i - . . _ _;_/,7_ , . ,J y,.; t . _

_. . _ j .,_ _ . . _.,_ . _..ge ._,._; . n =. i

-d _:~.g Y s .=-t: : = . .  :.r. .1 *-

_5..  :.= == =.~- f=t:- ____.2._i -t :=_ . f . . Li . :- t ;2 r j --* : l 50

. .'.dn .L =___p__=_=. L _..= .:  : =; ==

_.  : p=; ===r_.. m_ _:. - =.r-._____,_._ -

_=. _=_ . . h =.a=_ . = ::.=_-

. ._=u..=

=_u :- s- f:u ,: :-

2-

.. .: _y :

.m .m; :.l.;..,_:  :...

i-13: =lEiEr iiE==:hEi ! :: E=Es:L_= =t =fE:" "'= P $ D El E .F -W =P-lii=u *i@ i F=i~ .

"-i aff=!E=i;s -iE=E .. =ELW=i - iWi# =EisiliE:Ud:W_== iiEi =; 4.=Wn-W =1L-! i-I -l 4 ju c-F-MD:-a :=-d[":~==iF-tE:L=~- F = EE a= itki l tii=TIMi': +:f E fliui = -i = i = != l- !".-

. .. r =. . . _a t = L-- 2: : =.. ._ =. _-L= L: =. ,=. _.t.: n_ :=. . .j =. .._p. .:

, . u. . n=. =. _ .=:_.j 2. _=, .=.

._r/- .:.::: rI-

. _ . .  : l . c_ . I :

_ . l i

?= :(. .-P4Ei:TrdiE-I9 iri- 9.iO;I-iiG- /!.':Er;"r:4 uln i i F P'

.1 < =! I:M". .:t=. ... . ..

. :. t n _. . .. m. - . =. . .: _ _ ::...f . . .

. .. .m! . =_ . l =. t=. . .:= ._. . : 2. . a. . _= =

._.! _l_= =_. f W.. ...= =!._ .

...._l.....:.:.

..:_u_ . .u=. . s.r_ -...a.. .! .l. t

[2g ~ l 1 "PEl ;ji i=h : Mil"#'=if = iWi=" ".-r~5 F -!ruiE3E=l@':t AE_:Ms:EEi:R-W_-#_di;=E.i:ElSEF/ :{-J: MM

. '9!?/r=i: ,#f:_ #j P i! -1, i .F l i;=. &

J

_  :-J: 0-l=:.: w-3 a. _:.

.a  : =_p.x -=-u==y4. ==--== =a_== ==: =-

q  ;: - - .

, 3_ = t =: = =_-p.: u. . . - - --+=-_- = _:.= =.m a . :. .g:=:_=nt.a v_ e.. r:. l .::.:.

...-.=-+:.._!

y2 . .

4H t = "iifrE =: =_1.:,=T'M.I:Tri 2-;:!!=#7l'"5="=4=s .i=i2E' @#21t?=:5'#!!:-FE'= .l:

3. ' -E sE' E5= . i:ie.4 =4.. _ ,"-3:=e .W #=4 :51=Ei r__ m ..-T:1:;;~ f=*.il==u.~p nal 9: ~ = r- --

.. g ,o

: := y= =.- .=[.._.__.,; ::. . :_j. _ ;-._ . ./. :. - -- b_- f, ;. _.r._

p.. . 4:__.,. . =. =s-- - _ f. __ ::.f.: ...-:_l

-- . . P .r 3: .- - . l _ ,

. . 3. .= t e r. .- p.1-_ =:;.r.; .p.2 r / 4. :_  : .-r:r

... 4;. . _. l 19 l-:: .d.r- a i . i =- : - F i

1 ~

!  :!3 - i; et9  ;. af / t= i-r4 =cb. M -:FiM ! . -i: ' I 1 I I

- 1.o t7-  :! _:_CJM '.1i ~ -/ l i Effl'. L_" } : F4 :--h :l: 1J l- I e, l Q Eii..  ::=J. " n/ ; . .: r.: - -

. :..: [ ::: l.5.i:. n;.; l-  ! :- ,l l r

.l i _i5 !" J _-7  :.4 r gy 7;. . _ a ,-

~ -J.M a .r- - 2!.- . imi! .; i: j.- i l l .L"f.;:g P_::=_ a"Jg =W .

. :1qd. . .: y a. e,:qr_1 t_rgg : [. : : !. .;" l l , - 4=ii: :_: - /4 f ._;. kr.-};;. f. j ':n .J.~. !J . ub:.; F h :.l. .. l l 2.0 l

j . .. i - 4/ji- .:.:.P i e--l l- ,4 :_l p ) _t" 4- :::r : l =- I ' :- I  :

l l q -l /r.j r -i l.! l.! .

. i .' .ii l' ._.; I i i  !

I i Icr V- .

I ! ! -l- l 1. ! + i i l i i  !

l I2/i  ! I I I I ,

l 1 I i i l l I l I 10 '

l -

/i I '

I  !

I I ! l I i i l- i l I l /l I I l ! I i l l l ! l  :} I I I

/ l : I I I i  ! I i i I ! I i I i i o/6 O I ! j 10 0 %de I I 588 I i 489 t .t I i 548 F

(r De i

"T*e I i

$60 700

%,w % .

• c.,

• .  ; . 6 i , 1 1

l I -! l l l l l i l l l l l '

! ! l , l I l !I i -i l l l i I .

I! ! i i i l i l i l i i l i

1. -l i i i! Il i l"l-l 1 l l l l l l l .  ! .

l- l' L" . l- bl$ .hVam Ar,W "TL -ardd '

Iii  !

_.. i -l . _5- ll'- =I.=f!!.

~

ffr c.: f"'!~:- - -I'  ;.2 f : . _ l. cl .al l

. : . __: .= .

_ , . . -, _ . ..._.v.

._.ax._-2__. .; ;:

. . . t ai:.

. y ..

h . ..=_; .

4n.l.;

.-l .

l j. .; i

.(

l

. _ ; a_ _.

. _ t- _ _ ...

. L 25

. n . p .. ._i , .

=.:. .: - .--  : -

.:=l . ...rz=  :: - 7==_Jf

. =._ - - -==:_--

=:s.- ._ ;

_ _ . _ . _.. =_l:=1 =_ _ .r ===:-l ed .=.

_  :. _ . . _.3. = r : . . .- -:t _. . .....-(  :.= . ..; .

...:::+.  :. 4. :. :- :.: ._.. . .:= 7; : =- -_ ..

--t -=-*-#

' . __ r_ t- n. :. :. _ E= p. .. .i. .. f:. _n; .7_3l

. __ -"-":- ._ p. _- e,.,- p_ _%i:.

. ...L=._. p__.L .=&, _==__  ;: - r . _t._

.=.=,:.=_

40

_; .__.__. - - _.__ - -~ - ,._.

-- = -

...=._l._;_

- _; ._a  ;;y.__ . --PE::.--n .- - . . . _ _ . .._

. ___.._y_

\

__.__._= _ , _ . . . ____=p:.:_-_=.-_=--._= .__ _

=.._._:.!._.__._._ _ _ . . _ . _ . _- _=_=_._.m... _ - . . . . . _ -~.i=...!.._=__.. _=_. ._ = _;;  : . = ... _ _ . _ _ 1.__ 1

.;_=. --

.__p__.=:=_=-

= = _ g=%

--+ -- at-- r._,.a. =l=_=;.;.:.=.2 .-

n = _:: ; . . : _ __-

_ _ : =- i l1 1 ==- -..:.==:-=~---

l 3_ _ _- .

== r 2.-- - .:r-- .

FC.I c r. .;.=.==__L i= r;: :-- = = .._:.:  :. _

l f

C -- --

._.___.__.__._...__-..__.._._=_ .

d l

- * ,_-  :: : .: r =._:; ;_ .'=r_' . 4 L ' t= :_:.p  :; =  :

y ;n;;---*- - _-.---  !-,  ;;- =. c l

,-_~_:*- ;;;;.,,, .. /

.'~~~

._s

. . . _ *g

_.g_ =. . --"

._f_..,__}; -

5.. .. .__

__f n_N _ _ . _ _ . --

._1_.

E

. .-. _ . _.. . . _.__L..__..... _._w_.

6._

".=_.".1..".'-

6_

_..J:.:. ".=.t. . ._a.. . .. . . ,_, _ ;_ ~ - . _

L_

--~n~.-

= :.: s :- - , - _.a.__. - -

====-=:t -

r-- I _ . . i_, .= :-=:2 = ---- = t._ / =._-~-

.;r f. .- __ . -= ;. ;.- :r:. _ L~~~~;

Y - f -_,, , _-  ; .; __ .l _ . . =._-t.L. ..

- -__9. l.~3: _.__,

m _ q ._=_.g.;

;.g -

.,_2ur.,,_d~/g.. . . r.- e. . . .[,:n- ., .,,,

.g.,._

,, , f:E r =. l:, .:,,_,

.f - e. .::=-- . . _ . __i=- __o  : .._-J r=:

% #5 E !BI' .E.ZF ~"_"_G__ - =f_r:FiE -

@ # F ;~F D E'9 2

-_.._g_.- =_T=_ii=f 4". "g...

. fi3Eb ~ E-i=:E--=#E1rTr P

. -.-f .. .. ._. _- f ._...---'--_..f

_ _.l . . . _ .* b" . _J.*J. . :".: . _ I_"-I"_".*" 74."" j d'*"._"T. t :.r". *i. .:'. :*.":

. - . . . ..

• M. . f' ' . ""t* . 1 . h_" . f_r."

".74:7.I:'..

. r =-

( -l ' i> #-lENEi

_lE':  !#-"_ ~i.  :::@:i-i~5 N ~lI5)Id fi#YE N I5lfr9 J_ E  : "!I "l"l- l 10.1F.2 < I T 1- .!;l::=:-t" M '~p::n;;iiii _3_di).: gi:i:.2. iE-p.=_l@j;: ql....l :: e .l -i g_ j

1. .!E:l% =5 . tidi ._@ l l~ -l h _.l =.;

. :l _ .4. .. :d_i.t==..!?.:.3 . 7-i..!. .:

.. _. .i_ f:E p."...!!.

. .. i {r3I../ . _4. E_ h_i. [:-

. i". l _j i i

3 j  : i: T.:1 -l:.. l . - Qif.fl-CH e: i '.i.d : p-34l:"/.= .fli- E_V a -@. - l- :

, i j .;

L -- . " 'i ; - ; -j -K:.f ~:iti-hT: ,i:E@i~:j:= Ijj .r:. J =2.p_LK1.;.2 7 =E!Q :;t :::::r~. i JF ~ b p j 3 4 . . _ , .. . _ .

__F,....iv :C 'I_!'Ji - T*:l:nI::ie

. . .. _ l u.../+i : _ ,C. . _ ~

l ~ '_' afi 2 0 F. " 2 -l _E .= m --h.=i.I z-

[th . _ i . =_. __ j _. _. '.=j __-i_s-}

.a. l_~ . _ ~ '~iE3:Ep__.___i_=..f.=_ .. _ .,._ g . _i. 5

. . . . .I': . .4 .- .~2=3

-_ :_i #h .-F. .E_}..-:. :t wi:

l l_

. .l i, c I . = ,: =_- _._: = E: ==. . .= . :r.r . ,=_=- - - ___.=_;__ o- ~

._ ,s .

_ G- _ . ._. _ . _ -- _..:  :=_ )r :. . . - ==:_. c:=..___:r.=_: ._  ::c:A_ .n-

=='- .= h.tr.r- =.C n .:= _ .4' i .: _- -

., 3

- .i

-~ - r- - - -r . _. . . _ r-t--- _ -- - - -

___ - __ .____.i__=_==.=/.;;. bu ::==; =t: =

g.: .:==: .-/. : -

=. :f=. u =-

-=a. c.. .:;g=_ 1. n.,.. . .-.g_... _---r---

r_. _ j

$ .r.=  ;.: p

=- --- - * - - - -

-e x.. =_; . 4

= = , - = r e r. .= . .-  : : . i _ . ._ -

.2 -&_ .. r=_I  :.:- .-

. . _ _=. -- l : - : r . --+---r---

s. . :a. . l . .u - - v - -- - -

_ _._ _t= ;- . -- m=.

frt._ p=. ., ='II d , . . . . . . .. __  :  :: .f _=. . ._. .[. _I" f ~;#ir =T l : =-f. ..,_ .( _. . _.~. . . .._ :. __. _~-! . __. . t;--E-

i. '

• 2" ! - 'i -

e

. _ . . E_ .~ "' - : c:1 i i:PiWQU"5:

~

h "-d;It:5I'*'IW/- '.F F l[- j i ad it.-' ' "f f. ~ ~' i:- :h. - i' j " ' i ~ l 9 l W:]F si 2~ ". @ 5!M-@:23 /:.r ITira j ' -

l ~lif i--l"AI:fdf -i.-- .

-l 4

=g/,pii ;-

~

,  !  :.L 'i i-  : .r- typ;= :r 1 =. .:i:j.'e-- .--@ :)E7g:bi '.-sE gh . .i .~e. ;D i- l ., t.

4,

=

!  ::: tup : _.M.:} ; - -- ~

!-frip:J. _; .f . l . -id j.L _; .' si--l:f : r .:: . .t;- ( . i

.=l..I

" .:!..=. ;. - : '. lr

. . ..:: ;== .r. .. : a: ...).: 4 .. j '

l  :.._..a = .: - =-/::. . . n. _ p _. ;p. .r.= :_::.:. ru: - J.

M = g= j:i: pgj= ..A a_ .{. '; ' . -giz::: .} 4 .. / . .:.:=L-i:l:(;:h ..:.:i! _i. . p. .  ::f E'.j- l 6

i ,

-~

j.

,. /.

. :; . . ). ..; . r - l

- J ~ ;. L r. , : l... . : :=  :

...._.4.- u 9 p_E.: .=__ = .h.; -12 l

.0_+ .$. . . . . . -l .

f 1.

l l .

[ ='5 #. f:.: f. f .l 'h h C*

. ' , ki[ .;(

• f. f

• l l ; .i-l fl 3.;h.. p : 1 I t l -j" -

lin -

l j l  !

I 1 l ! l' / h .

i :"l:1 I l I i- 4 f T"' I  ! I I i ! i ! V l .1 l I. I I ! I i  ?  ! I I I I I I /l I i l l C '

!' i l I  ! I I I !/! l l l l l i I I ! l 1 I I I I VI , i; I i i l i  ! I i I ; i  ! i I ! l I iI i l l ! l ! l l ! j i i 1 I i

& */ l .t 4 .k .9 Lo VL M M l8

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - . - %m C~O

l '

i

• i l t l l l 8 t

l i l 1 l [ -l l ;;i:y.l t. l* i i l i i i i j l I  : ' i I i

l }-l:7 .: l, j xj' ].: .l. , ul  !

! l t j t l l j l l l l

_ l L - l. U:- . i i.fi.h. f E_L~ .:.  : lJ _f. :. _ E. .l . i. _. . L.. . . .l  ! l ; l .

j' i .l  !

.l. '

l ~-l . I'-! EUd"5_:.  !.:E- i::f: .i)=-!=_l=ki~~l2:di-Idj :4 T!;- F--I :- l i I1 Tl l -~ ~j..l-'~i l -5J-I::~"=iEWh=. '=E= l ~:-I~~.::2.I=_i e d i P S~"5:~l' I M E N -I E @ - M 2:t;E - S'- l" h. ' .-:

l.___ _l _ _ -- ; ~.~ Z

-'--~_--;_-,'3===' =',.___j @. -l1l i-ii ;E=-i .[.:-i_ jj@:[_ : "i_9hj:'_::;ig

.__.p.__

_._=___===:-

4 l=_ _t..-==4,--

_; +. :::= --- - _ _ . _.:4._ :.- -  :- _ ___-

- ~ . = .: . .. .N .- en == t= =_ =T.=_ : == ,t = m- -- -. n= -_F__=_ _o

.=_= , _g ;_ .;. .._

l=-::--d .  ; _ - - _ _.

-_.._._.__c- .-_;-.,=.__;4.__;.___..

1

._~ _.

- i. = : =: =_ =__: _ . _ = _ _ .--

=

_ .--_ I. _3i=~ _ Eh_ i =.=. .__ . '~_=. __

_ L r2 :: . _-- _ . ._

._ i=_;_TE ,J_?. _ _ _

l f =:-l:_9k _ . .__

-_._r- _. _:-

u_. - ____c_(.. _ _ _ . _ _ _.___..._2. -. _ _ _ _ _ ._ 3_. .__.: . _ .; _ =.- _.L- =. =_ _ _ _. ._ .._

. .____s - _~ .

l. .__

._._.__!-=-__..- _ :r- - _. _ ...:____;.--=_=-_ _

_ . _._.. =_ .=\i__. __ . _ _.

i .-__.-.____.___.____.1 _ . _ _

l r, ._

\ .__. -

.n e'*

.::=_--- . t. .: _ :: _ .: . .. -

• 7 _ %. __. _ . . _ _ _ . _ . --t=._=.. -:_-

1 -

r- - __

e - _ ; . _ . - _ __ A u__;. __ __ _ , _ _ _ _ ,

_ . . - .-.:_..______-._____4==___.=

_s=__.__:___.__ ._ . T p _,_ .__.__ ,

._____i__.___._i.__=.._.._.___._..

• ._ _____.__T-"_

_ e _ -~-

l . __ _ . _. :r- __ _ = _ih - -+= e ---f __.____l=._.. _.__ _ _____..--4+3

.= =

_.__..__.;._.__.~~

.A_ . , _, _ . _

, _: ,1l .: _ _ _m

. _ _. _ _ . - F. A ~-'-'..H-.-~-"'_--"~'~' +_ -

• =:r___..-...

=.:._

-: _ . .. .r. .--. . t__ . _ _= _2 =. . h. .t._._____.__m_____:_ _..= - __ ._. _=t L-T=:5. _::r1  : 2.  : .{.. _._  := ..l' e .-- r:. _._ :l__ .: ._

,$ r.. _. =.; .NN -: -;-  ; = - - - -

c.,::.=  : --

=.- _ _.4 =2: _; = U

- rt.--=== - - -l. :=u- -t:. :t -

. . = === = .:u r-==t-

_ _:._- r ut __r _.:. -- :. :.  :: .:-_- sl: ..:;;;  :-.:-_...__-- .

'~~-'

Y-Y 5:*:-t\ Yf $=T---Y? ?k b' -- * - f~~ I =' ~I1 N 'E$ 5hI?:~~b?

7 E -i 3i=t=_=Eil M _ ._ t=E5tW=f=+.=; == l ==" =!"E  :=-lW-N =le =i 'W .I n=-

% _r---- --h5:t =A=3=iiEWEM =:1_E-;== ' E!Et3i:"= il:.dEU=b=" '=9.=:

=d= 1 -!h! -L- o r[i L...,...._.}...

j... -

- . . - -' :U. :h =.3. ._-. h.: . ":

__ . _. _ . . :l .: _. . :_._._ __._..  : r.

. . [ . .' l." . _ . h. .r. h : ' . l _._. .

} . _.. .. . .t._.

] .

.. . .-. . _ _ . l.a. _. .. .:.:  :, :t .M .( r .. - -.

i. -

C Gm a , r.n- a =_ g=_ ._ r :.::-p m.-

m=-,

rt- ..j .=rer - _\ .u.:=t".--- m t_ .; -

.:.? ; __ *--t. . .z = , = :2.i _: -=r 2 j. .=_ t f . .re i

?- ?~ . .I - . I

-t-r -.._

.r -

L.

i- . ' u,. _ : .: =.: .-. =a =r=:2.:= = .j : :2 - --: .-t--:: _j. _ . .:. .:=. A r

__\. -:

-: :. t=_ _ _.__ s _ . =I. ..__ ._ . . .y _ . _ .__ _ . . . .-}: ; .- . yg --3. -

- ,n..-u _-

i 4 , =-. _r . _ _ _ _

e

= 4. .:_~=_ t= _= - -

_=t===N. z; :. .. _, ._=_. #. . .b= _=m l =_:r: =__ r== r. _._:= = .. 4: l ==:;=-

.._=: ._ _ l  :=. - r.i---.

._, . a . _...= ..

r

- [=:, 3_._ g. . =: _

7 5..s_. 4u .4._=. ==+ =d ==== ~- -: '

=4 _=_ i_._ . .. =_c>=_ .  : h . d :._=_s -

p, _=_ ;__._____._=- .._w e . . = -- =_,

i=._

m: I =rw _. ._

__...__.- A. -

._:._= __.- 3t_o - -__- _; ==.:=3. = _.: _._-:n-- a : ;_..-

ys g.t__ f rw -: _. -.

_- - - - _=.q- -

--==r= -_. -

==2 - - -

m.. -_p .:. . _;. __ . =. 3.=. . g=_ j . .x. _ _. _ . . - .

2 _

_  :.___  := ;V == g .=n :_.:-- = ,j n=r r. = 2p r: n :n:,.-

s .

l . . :.a F_ :-. f.3.  : - : ==: :. t.- __ , _ . . _ _ ;_ __. . _ . ..__.s . _ _ _  :--t_ _. -.

__.__.= . "_.a _-_. _

p g a g. y - . _ . .__

__.._-._ _. _ = . . = - _ ._._..p...___.r_=..,.

___.=_c _-- _ ._- _ _ _ _.......l__._.r.

=.:=_ _ _ . : . . = . _ . =,.. . _ .. . . :. :

~ = _ . __-

=- - =u=  ; - .c = ,;_

a.z._:_ .l i- . _=z=_.= :..z-

_.. r.- ==4=====:

-- ===== r~.=_r=. ,_: = -l=:=m.. = %. =.c - _. - - r -- -- =.m . i=.=. - -.

!  ! - J -G #E. DIED ==E m -Z=~E=.i_i- ~ ii == e= -l -i=E=_%=i n M EE'3=t r  % = 1=i k -v ;w .

=

T.  ! V i i =i _~~_I.. _ r-

_. _"._9 4s1Es E=_;EE t=:

=_ 4=?_=i "_.*i:! -x !  ?-.

m w '

I = .r.5: =

_==-- =2 r =. . _ . . . .=i == __ =_= :: l w. . . ,:!:=

5!=P",- -

. _ #* = t _.__ _:I4=N f=AcL =1 _:"_

- - ! m "#..= + = r! s_.

f'1 j~""~%. . : =. " r- ; =._.:=..=- . _ . _ . _ -."- i . :=

2.:.t. : e -- - .- - . . n- . . .; n=,

2 = L = --- : _::_ __ - r- - : := .= ,=_. L_ . __.N.. - - -i==. ._ 2 = uN  ! -u- -- .l:. _r- .: :4 _

.a I _ . - -- .

d F

E :s F i.E : ai== 24W= :== F=t= . E-l -d= E',_-i' :E M :ibl=1fl u:. I i- r-i 6 :.: -smE:_+mi=: := t= isi=y=g=-f.:: 4:"-g g= =4=w_ g.-g.3 =:t 3.N 1. 1l - :_

l p u ,.:::p ,--g p s =r= =-l 3-- t . . . y: ==p4=_rgr;.1.j N : :: =! -

i l r_ q i -q. .i . -j r-. 2r._ .:c.g- ._l . p l 9 .-r=_; . j =.g.- -

1-N  !

j l ;_. z.i . t , , ., p. L .i- I. :24.M l . j, p .a j j--gi . lrlN!

l t ! . } 1..l. j. 1 Ll . l .. .j . I i I  !. - ..j : l: !. j .i - l l N

-l  !. . .I i i I I i l l I i l "I 4 i i l i i N I >

! i I i. l l  ! I l I I lti i I ! l ! I l \

l I  : ! I i l I I i 1 l.; I i i i  ! ,

i i ta i I  ;

i 1 .

l I l l l ,

[ l i I i j i i t i i ! !  !  !

O q N N.

3 y ye , 6-%..re~0 wm

% 1 I D,,,ISP,.O.S.

ITI.ON..

u . FORM acrencacc on omcc swooi.

,, - i i. 1 co.

s a ccr i

l RSD MCMilment of Equipment Roau 3152 Roof Scuttle (Hatch) -

l fo FRON oATE cW11 Files c'hnirman, RSD 1 Demanhar 1968 FELTY /jrf/51290 I concur with the analysis provided herein ani fini that planamant of the roof scuttle (batch) in room 3152 in its new location poses no incmased ,

hamrd for persnnm1 who wouhi use this hatch during a reactor emergency. l c

- ms Chairman, RSD Reactor Facility Dimotor l

l I

I (I

^

D, IS.P. ,O,.SITI.ON..

. + . -FORM ,,,, u atmoca on omes svueoi. susa ct RSER Movement of Equipment Room 3152 Roof Souttle (Hatch) 70 . PaoM Daft CMT1 Maj Felty SPC Cartwright 22 10/ 88 ROS SRO Cartwright, Munno

1. The roof scuttle for Rom 3152 is to be movai to a new location within Room 3152 (see Enclosuze 1). 'Ihis report discusses the impact of the relocation with m43 to radiation doses recievei during an amiriant.

2. 'Ibe information on which this sunnary is hu==i is provided in three enclosures:

a. Enclosure 1: Roof plan drawings fr a contractor.

b. - Enclosure 2: IF "MIGA Emergamiaa - Dose Estimates", dtd 20 SEP 78

c. Enclosure 3: Drawing of ooze in Northmost position (231) with does estimates written in.

3. Does estimates for MIGA emergiances were determined by Mr. J. Arras (see Enclosure 2). 'lhis summary was reviewed to determine if the relocation of the roof scuttle within Rom 3152 wouLi significantly affect expected doses to you using the souttas during a resator emergency involving ospiste loss of nnnlant, with the core loosted sa the northern most partica of the pool.

4. Rom 3152 is esperated from the reactw deck, therefore any fission products zulaamari during an amidant wouki enter the roca through leakage. The does contribution from fission product relaman is ley.eut of the location of the roof scuttle.

D. 'Ibe dizeot view dose rates were calculated fe points withir. the reactor room (see page 3 of Rnnlanure 2 of the Import). 'Ibo position of the the ooze for these estimates was 906, z the Southern most portion of the pool. The l I~ points salmoted at nail 4a8 2mvel (the same height as the roof batch) were ,

directly above the core, ami the furthest point from the core while still in '

direct view of the core. 'Jhare is no point determinei within the equipment room that allows a dizeot path fa gamma radiation from the oore to the roof. The path from the oore to the nearest point in the equipment roca ocntains a =in4==

of five feet of occarote (estimatei from drawings). Tc reduos the intensity of i

l I the gassa radiation to one tenth of its origiani value, at least 18" of oonorete is zequized (Radiological Health Manr9hrv*, pg.149). 'Ibe five feet of ococrete wouhi provide a zeduction of intensity by myphtely 1000. 'Iberefore the

{

i does due to games radiatim at the na4 ling inside the equipment roce is I

. negligible, ani the reloostion of the roof batch wi2.1 not increase the dose received from gamma radiation fr a the oozo.

6. In the interest of ar=1ateness, does rate estimates veze performed with the ooze at position 231, or the northern most position of the pool where dizeot view of the ooze might be P4hla frca This I requizei a new drawing (see enclosure 3)an azzi parts of the Equipment assumption Roca roof.

that does estimates results from enclosuze 2 wouLi be synestrio. 'Ibe new drawing shows that oorcrete attenuation of the direct view of gassa radiation at the existing roof

-~ _ _ _ __ ___

( 2) scuttle and at the new roof scuttle hetalled at a right angle to the com in line with the old souttle is over 12 feet. C1.oes to half of the roof h ro m 3152 would be ahnemwi from dinot view of the com by over five feet of .

oonctwte. 'Iberefore, the radiation does attributable to direct view gama would '

, be Approximately 0.0 R/hr amose half the roof.

7. De scatter, w skyshi.ne, at the roof would drop to a working Isvel due to ,

ettanuation of the p oca by the roof material. This skyshine level would be i less than 0.1 R/hr on the back quarter of the roof. i

8. The planamant of the roof soutle at any point arouni the back quarter of i the roof will not change any previously pmdicted levels during Wdant I conditions.

C. Enclosures Philip . Cartwright as SFC, USA Senior Reactu Operator

'IO: RFD FR3(: ROS Ott 2 Mark Moom MAJ Palty Based on this analysis, l depicted b enclosure 1)planament of the roof presents no increased scuttle hazard forin the newwho personnel location (as will use the scuttle in the event of a reactor mergency. l l 1' i

l

/ 6 .

) '

l )

I/

ames R. ty Major, USA l ROS ll

1. 3 E,.g <

.[ .5 g .

,7gi 0hp 5N .i n gyo '

e{" . d l ,) y* b aIE2 . 7jg Lh(

- [

$ @ ;qk=0 f E ; ,

G OEL'0 !I j4bI @E0 d Ohy N

I H

? ja n A.

t g

ore"3' $; pbg w4 j a [n y ,o0 } ,

F .

i ~ s t

P

, a

\ + h

• o iV b'

i 1

l '

O -

o R

P G

j H

T N

h H .

- 3 i I w *T, t i H W e.o i A e e.O. S) H T t P AM CG E' nN = A L T ic o c PI N G PIC E I.

O A1N ul.l _ .

s L N nT -

DN F' =

A. 2 n t

AA I .

e TI ol P t

e sTT I

NA H eA T Ni o e

E T. S TL T u AtT Yo e. o e RM mN n E N s 7 Ve A NS U c s iD o D L O O s EO A- .

TL A t s w

G P C oN 3 N, MT M N

i T

I C TI A oL CA H eE ^ t NR E D c T P o

D2 Pt UT S t vP T i

v i a A 4

iE T

3P i .e Dcu o. Wo *e N D

AO L c A rS e o a sn T L i v4R A ND P E N eA =

u~

cA D o o

P x5 o.I NL TM C C .

'd EA , rod PN ERI *o-P. A I

EE

.R EA -

E T s

rG e m

o o

n m S r N. P u- i s 5 SA '3 DT CP E oIN n s c

N f - 'A AP S _ i i A

V E oS o4 no- oH EiAt BE r I

T i

/

PTiP d OD AH t

P. T oA sS2 s si T

n o O

I

(

I PEI W (R 2 *-

RN Et GT C cH o

G_O A T e x o w T

/ oP E1 PE Rw t R a A

( G N .

w P E GU o uP N ND O s a

e T o C i

t "

t P s o S "l m t '^ssr r tii I

lH n I S o H EE r w i

N f C A - T l r

I T e P' D- U - u -

F SPID ie L*. D 7.

r '8

- i sT iE iAo x s s s e e

/

vT G X *^ x I oo LO e2 ED EBN I -

XsO EAM RN AN T

DN NI F T T

n m s D

- - F, sL AH E7 H o o - O N

2 i S C G r ro e.

=scrrn MF G . tuA NA MW I

_ u r

=_

o 4 So oT AL EE R _

s H

t TN n E FF FH ,' >

t u S C AD oR U L

A Ot DiA e. R Ne iT

. f w e

n mo T, r O

P GT cE T - N = O E iP sN .

io R v GMe I

E T L H

IP T noS i U xO eC T

3 o i wno R L

sM rA tx es .

NA OUW i

x "E r- _

a r T x m A CF Ef c T

) 3 _ os t

2 GO

(

3A ]

em 4 NH i ne TT = . G SG e IX H '

/ I

q , j , f( '

v N Et l I

o I L m D K L L

r t u X "=-

I ',e > s&

= .-

_ K pl uV e n L U>

Ai L

I U

CP e S sU .

g ,5 o & 3 u_ D oz3ga3(34a_agea

%/ /

!ir.YO1" < - ogu Eg Iga O n .34O f 0 o .=t'J 4 ; fjf

,g5 7 o.

g o i2st l

. I l

l s'

s _ .o ne

+'A

• V

$$fkbbb

@Ic 3 YhI (s

!58Nb wn3 >e n ;,

r 9  %* '  ! P;6.W > S INA )

.n ,'

\

cin y  ! I 4 -b s~y O  !

'l ,Ay ai

\ '.$ ,

,\ ./

/

y c, ;-

p a As a P

-4 l l b V, dgOisi!

t s x f$

VQ \

mp  !

$ jij FS }nyl.s71 *3 *

["q J- dp 3g!a!ii c ,!

9 le i.

'!57 T

g iI3 5A ey l  !

I la a { { l l,. P fA ;i 1 U'

l '

$d j -

I dS tl gm "

...eo i s g aclenoe i> t l 9 .

IbN I 3to taE ~c et G

I 1 j If M

$f i 2 3 l l .n 83

.. D B5VG ^.~ X - 2 a y, oc

--. T E Ta u^ L*MIA*"."" M*e?-

I =g upame urowemar:cmon.SYSTBS

_ _UPGPAP.CRooP_5: BUD. DIN.GZ4 2A__.. -

E A I4CRRAN,STElPHCMSON ~ MODI f*1CATIO t4 l P 000 Z i_. _ __i_'_~ '_ . _

l

.t8 A2.4 A LumtavM INSacT ScMEEN IN EMTm.vDE O ALvMINum f*RAmt.

I l

1

-New IG GA. GALVANIEED METAL  !

V E NTI L A TO M 7 '/2.* A 11 * ( v e ns m y # N PIELD ') To muPLAcu. mAisyins XEt*T8 OTO N . ..

., 5'W ALANT (P EO. S F'24. T T. '5- 2.Soo) 4 ,

i y

21 G A. GALV. METAL. PL AS H tNC, j 3

(' S E ALE D OCRNEAS)

?.

-I <

II ';>

DMS PLASHWG [ EMTSNo G'

? / MimiIMvM CN maora MuMeit.Ame ')

e.'AsTomaasc. xA,3Te I .().

4 Q

l .

me.T t.etANcE

=veronT s

s

,y ^ \

N e'

~RF. PLACE. P.OOFtMG AND I N D U LAT1oM

- ~, . ,

\ TP.EATCP WOOD BLOCKING A I .New Gciose_ Weck VewriLATos w. .cAct I 74 o r t" No~.1 T -

_PPoviDE NEW SLOCKJNG, CART, DASE FL ASWING, coMPoblTioN PLAS t-hNo,

%t4D.MSTAL- P'ASHiHG SiM* * 'To A* S d** N Aeovs F ALL. Duct 5 I

Tvgenu, ssowN oH THE I"'AN. NDiP'T HEtGH73 AS dr.q**uingo,

_ P'A>4 SM AR BE 1%AtSED Mi>4tMUM'. of 8" To ACCOMcDATE NEW PLASN'r%.

1 canasaac DETAIL bTO,2 ormarn. cur op vie am,v an.t ,mes.ives ca scio I , ,. # WN b. I'* ~ D HENRY AEMAES, NC.

CHESAPEAKE WAlpeesf0m. O C DIVISION ARMED FORCES RADIOSIOLous R :"S*RCH NSTITUTE NATIONAL NAVAL MEDICAL CENTER BEThCSDA .-MD.

I I consumns Enskwes

a. m m er., m aryi no L.PGRADE f4KNANCAUELECTRCAL b'fSTEMS UPGRADC Pso o F - B UIt olN G 42A '- '

. cocHvw,3TrPHrN scx4 mod 1 FlC ATIO N Pooo 21 1uD poNKrsuceT--

1. ~

ARCHITECTS </-e,

- - - - ----.-.-..->~-..ei.~::-

. q . _ _ -

i i 1 l- S" - h 7_ w ITewH PeAo

. ram-c.

t.A_N G E

( 4 id) r.w sTlHG v e"T r"' Pe-i  ; ,

l ELASTC ME N C MASTICen N  :

l i

% mr. o te_ o eos 10 se,.

4- l  % JAtmeAANe. ( ANcso n. To

/ -

-'T me ate.o'wooo N At t EM

-l  !

. VENT PMe. PLAsWiWe DeTAtL Mo. 3 extermo r' ire.,os cowouiT

. -~ se At.^HT

_"' s.TNnt.ess sreet et.Amems eANo

_._.; 4 :s t.sio e...z u r. o + 3 1

5 i New OAst Pt.AsstNG [s eALao)

%p

• m t. A 3 tom *Puc- M ST*

4 s . .. .f'.

I -

x -

U g .,

_.v -,..r :rn _ = - -- -, . - - - . : .. .

ir , .,

g

'I

_]TefcAL CoNoulT Os BetTCA5Iilnh - DETAt L No. 4 l

I.

""*""""**'"***'""'***'"' l DETA1L do.'S Q4 """*"' CHESAPE AKEDIVISION I~ I pgWG, K.4 wansamatom, o e ARMED FORCES RADIOS 40 LOGY RESEARCH NSTmJTE y m g' NATIONAL NAVAL MEDICAL CENTER BETHESOA , MD.

LPGtADE NifsCAL SYSTEMS I con umn, sne n amamore.: w U PGR A D E ROOF - DVilDJN_G_.~41'A --- -- -

!- Co cHPM, STEPHEN 504 MoDIFic ATtoN Poco 2\

N g$cg g CET T -- CCNSMUCTictLCcNTFACT .No. N62477-s t- C o t 52. :

ar +v NOTE: FEMove EFtSTING CoNooiT3,e.TC . AND REtNSTALL AS DtAECTED. i j

New .oso ALvN\non corsHG " \' \

Pmovice. FiLLeg, wHejit,3, Paquigro pes DAAiN ASLE GUTTE1't. WITH COMf' AE55icN '  !

lea r'r P ic A L A LL Jo eN TS %

g suppog7 e \ c.1 ""

=

Z t MEW Cof tNG 4 i 1 e, c _

c  ;

[k .N'h9MhM.g g l j -=

g I .rI o

1:

I-

==

==n

{

i

{

,z , r k

l" g+

\ i N ,

l

- e"> Tins ~r^' orNG .

t l

,,,,v,,_.,

m.L Au. a . ,~ T. i'.w e s,,T. 2 l

, tu

• /

l - ) .>

I  ; . \

j p  % l  !

j crus .

g(/

3 h  !/

I =NSTING PARAPET ' ,

r,m - v 7 _ALus.H.L o,-s ec A i

1 m ac New As.vM. c.% ion et.4Te.(.oro

/

/

e = = r. ra s A r a r r N s s

.I i j

'TN'8".New) ove % Jo'NTS ot" E* *T' cct'"*

PAca(snau,isPet.Aca) g 1).ETAIL No. 5 : NEw METAL CoPmG INSTAU.ED ON scAi_ . -st i to" ALL. CXISTib4G ME.TAL COPING New coring sNAL.L*De. .aTTLE. SL Coring AS MANvPACTvPa.m D ey con symucyi os see.c ALTie s , INc. c or N a, S H A s.L. e s. 17 ' w ioT H , ,

. o so

• TNicw.Ne ss . stucco ALvMtNum MNISHED IN K.YNA% 850 0 I

g STAN D Ast.p coLo g .ccAriNG S. Coping To eg tNsTALLED WITH evTT Jo t N TS fl 2.'-o* O. C. W I T H A 3/g' MiNIMU% A LLoWANCE. Ptwo. E.xPAN5lo N.

. tNTe nNAi_ Fx u. uNo -vr SFu ch To on. PsovlDSO AT Au @tNTS. ~

l

' TN STAL.L.ATIO N 3H ALL E'E:. Accomplis Hm.cp wrTH 6' wiet p.ETAIN EPL. i 1 C.LI PS AN D Ques. DVfECMETEA~ COM pm.ESaloN PAQs PbASTEMSD AT A LL sjolNTS.

.I Q_* w tDE. AbuMLNVM: HOLD CovvM PLAT E. AND DvaL DvFloMETEN PADS ~

. seAu. :maussTAucsD AT G'o' cmHTam.s av us e. or scnawS ca NYLON ComPM.s e$ noN PADS TO OS AUGN tO I'TAPt'lTS}^7 S.A~30M CoNDlTION . ,

. AND /ARCHANICALLT SECVf*.SD pat ofA TO Co f=iN G BalN G SN A PPS.O IN P Acte. . i

^ C OPING cop.HtfAS*.AND .NEf%TicAL AMe AS SHALL Eb5 MlTEARO AN D . CONT'NuoJSL'(

."""*'WELps.p .PAloft. .To .I N ?L AMT . P'I NI S N IN G. .

l I == mat w w i=v .=ai. =ca.ives p <.-5 CHESAPEAKE

.,,,,,,,,,e OlVISION ARMED FORCES RADIOBIOLOGY RESEARCH NSTITUTE -

HENRY AnaM, NC. NATIONAL NAVAL MEDICAL CENTER BETHESDA . MO.

! canas*w sarneers LPmADE WCHANCAUELECTRICAL SYSTEMS

! ses===. W U PG f%DF KOOF - BUI L DIN G .4 2 A CoCHPANpTCPHENSON /AODI FICATI o N P o o o 2.\

- A N D DO N /.E AVo ET -

..m...,.. CONSTRUCTION CCNTFN.'.T NO. NG7A77-61-C-0152.

i

l I

• * " M* $

l m

)

MINI,WUM 30' SQ UAAs. 4. l b. LE AD PLASHIN G

• Se.'T' ON FI Nt *J H ED RoomiN G P8 E LTS nN MASTIC . PAIME TC P SVAPA35. DeFCAE S TAI PPIN G.

M tMb84A N E. PLiuLS META 6 FL AS H'NG AN D PLASH -IN P6IES : MTEN D l ympg Pm. CLAMPtNG P%tNG.

ST RI PP I N G P"#LTS 1 EATEND G' DBYOND SCGEIL OF P 6 Ass N s ssesT.

l I

DTRIPP1NG f"ELTS j i

i FELLOC ATE.O ' Dm A t N  !

Ns  !

N 3e e

\,  %~

k\ CEACP FLASHING l "

ce )

Y q a F fli

{}

Pooviom com clamp pos - -

1. }g $

I

..O M At N (AS F%EQVifNED) h i

. m Y 4

i ,

-% } N M

I LTAPt.R IN5ubATioH 24" To omis

. PP.o v l D C. APP Mo >. Cf ' STEEL. f*l PING To e ^' 5T5NG V ERT' cA' P8 Pe.

. (iN su L ATE. AW HORf Z.oNTAL PEVNS )

.Prt. ovide. KLCFa' IN axi STING Ft.co ring P R.o M se. cannam 1o omals.

DETAll NO.Ti : RE. LOCATION OF Ext STING l EdOF DRhod ~. .c.

i l

i I ,j De T AI L. N O. G D RW G - X~6 osmafwtisT or 7is suwv aussu. paca.ifss aseesecneie coumase CHESAPE AKE OlVISION wasseserom. o c.

ARMED FORCES RADeOS60 LOGY RESEARCH INSTITUTE i I HENRY AIMMS. NC.

consueme sm. W s nan n NATIONAL NAVAL MEDICAL CENTER SETHESOA , MO. -

LPGRADE hECHANCAUELECTHCAL SYSTEMS i i

UPGRADE PsooF - but LOtN G 4 2 A , ,

Coc4FAH, STEP HENSDN ' M O Dt F I C ATIO N POCO 21 l j AND 00NKSVCy '  % q, g g ggggg l l

FJ r

l 1

g*; g V'

l PAovlD2, NEVV 8 Bit.Co PocP SCUTTLE, ,

1b6" A SL c 14 GA. GALV. 3 TYP t. ."5- 20,

• I1 3 M,to CMiea PNMEA (PAsNT 1 cc ATs ) .%

4 EwACT LocATio4 Tc DR DETERMnNED  ;

my CW N apL . Vt.  ;

\

. ,[

7%CVIDt. .T AEATED YvocD Bloc W NG \\

~.AND CAMTS CoMPLETEL7 AP'4VND h, ,.

i n

Pa ~

y roa Ts

=

1 .

y

- i_ X v y -

l PROVIDE. BAS al. ^AND CoMPU5tTI N FLASHING (EATEMO l - c" ')

iNOTE: P AovlDE (4) 3*A S*A! V16' STEEL ANGLES AND I 'M AITXh* .Spaays C. ON

z

._ :DiMtbA PN To cont-r EstlSTING..NA-[t_H.AN

.om. . . , o E r~ D. u e,A5..SMovvN

n.e ere. =*istiNG s r==u. 'Ao DerN. -

DEili EWo.7 i New Roor 5curTLt koTE No. I 1 REMbvE E.EISTING P.8CF H ATcd AND Cf oot. cPeNmG As sHoWN O N CON T P.ACf D NWG . S* l 6 D E TA I L *A , #

I j _. .r. , .nePAtm. j ALL mooptNo As maquim,go To Mats:w existiss.

_No f E. No . 2 1 Ew!9pNG RocP' S cv7TLE. To P .ENsN .

. . . . . _ . . .. mvio E N aw oLoc. AiN o , CANT,'5 A s E F4 A S4iMG y,.~. ;: I - 1.;

AMD..'cA%PomTioN rL AS HlRG As f*EN TMS DETAIL.

I

.DE. Tall. I'l O. 7 OEMaf astaf air fut mairf samme. PAC 4.ITitt EmeesttRose CohnsA880 CHESAPEAKE DIVISION DRWG.X-7 wasemoe7on. O c l'

~~

ARMED FORCES RADIOS 40 LOGY RESEARCH NSTITUTE I HENRY ACMM5, INC.

conumine snor==

sammam M=yi-us NATIONAL NAVAL MEDICAL CENTER SETHESDA MD.

LPGRADE MKNANCALJELECTFUCAL SYSTEMS 4

_ UP.GfMRC PsOOP - DUl LD14 G 42A CC CH rat 4,STEPH!'N30t4 MO DI FIC%T io t4 P o co '21 - l t eMD 99%%V@BT " -- - - -. _ _ - ~ - - _ . -- l

-2 - .. ---

"DIOP@OITION FORM

' Po, see ed Als ferme ese Am 34611, se pressment eveney to TAGCEM.

sePGASMCE ca OpriCt ums0l, W 0 t CT l

SAHP TRIGA Emergencies - Dose Estimates To SAF - '"0" SAHP oATE *'

20 Sep 78 Mr Arras /bsm/50351

1. This is a summary of expected sources and levels of external and internal radiation dose, in the event of a TRIGA reactor emergency. Exposures are es timated. where applicable, for boch restricted areas and nearby unres tric ted areas. Most estimates, as noted, are' for the maximum credible incident, which is, for the AFRRI-TRIGA, complete loss of pool water; i.e., coolant.

2. The information on which this summary is based is provided in four enclosures:

a. Euclosure 1: Excerpts from Final Safeguards Report.

I, b. Enclosure 2: Gassna Exposure Race (Loss-of-Coolant).

c. Enclosure 3: Fission Product Release (Loss-of-Coolant).

d. Enclosure 4: TRIGA Emergency Rules-of-Thumb.

Each enclosure is expected to stand alone, and may be attached to any appropriate document.

3. Maximum gasma levels for the reactor area, excerpted frorm Table 2-2 Enclosure '

2, are (approximately):

a. At the top of the pool,145 R/h,

b. On the reactor building roof, 36 R/h,

c. Outside the reactor building,1.6 R/h, and

d. In the centrol room, 0.6 R/h.

4. Maximum unrestricted area doses from released fission products, extracted from Table 3-4, Enclosure 3, are (approximately):

a. 1.4 r e/ day to total body, and l .

b. 3.7 res/dsy to thyroid.

I; e

4 Enclosures JOHN M. AREAS

- as Head, Operational Health Physics Division I

D A ';?. 2 4 9 6' """*"***"""****"*""-

l " ' - " " - " " ' ' "

Ew lo sac e c2

1 I

I Excerpts from: AFRRI-TRICA Final Safeguards Report Revised, March 1962, Chapter VI: " Hazards Analysis"

1. General.

a. Only that information deemed relevant, to the purpose of this study, has been extracted from the Final Safeguards Report. The evaluations and conclusions stated is this enclosure, unlike those presented in subsequent enclosures, are chose presented directly in the Final Safeguards Report (FSR).

b. Specific hazards have been evaluated and classified in three hazards categories; i.e., these for which one of the following statements is considered valid, according to the FSR:

(1) Physical reactor parmaters and/or interlocks eliminate signif-icant hazard to workers or the general public. Code I.

(2) Administrative procedures are required and followed. A somewhat hazardous condition is possible, but constraints, in the physical system, prevent any serious hazard. Code II.

(3) There is a possibility of a serious hazard. Any such hazard will be discussed later in this report, in detail. Code III.

2. Suannary of Hazards Hazard Considered in FSR Code (see paragraph 1.b.)

I Improper fuel loading I Variation in excess radioactivity I Malfunction of experiaancs II

, Reactivity changes I Loss of coolant III Argon activation I Fuel element cladding failure II Atmospheric releases II Soil activation I

! 3. Loss-of-Coolant.

l' I a. Assuming,v' ery conservatively, a virtually immediate water loss (" impossible as per FSR), convection would s till adequately remove the af ter-heat, after either (1) continuous operation at 250 kW or 20 minutes operation at 1 MW.

I jl The original assumption is based on aluminum cladding, and the current (stainless steel) cladding should accesplish this at least as well.

b. n e maxinma release in the event of a cladding failure is estimated Ii as: (1) 5.1 Ci of iodine isotopes, (2) 5.0 Ci of zenon isotopes, and (3) 2.9 Ci of krypton isotopes.

~

Enclosure 1

w _ ,_

i c.

The FSR does not provide direct information on fission product inventory or dose rates related to loss-of-coolant. Information given in subsequent enclosures is extrapolated from information in the FSR and other references.

d. Loss of moderator, caused by loss-of-coolant, would cause a reactor shutdown, and would not result in any criticality incident. Therefore, no 1

significant neutron dose is expected as a result of this postulated incident.

{

l I

l 1

k l

Y i

l \

l i l

3 j

2 I

1

m w_

_ _ ___ y _, ~

GAMMA EXPOSURI RATES: Loss-of-Coolant I

I-

1. General.

a.

The Final Safeguards Report (reference 2.a) does not provide direct information applicable to exposure races, resulting from loss of the TRIGA I

reactor pool water, and applicable to typical TRIGA operations. None of the other references obtained provided specific information applicable to such l- infrequent, short-tera, operations.

fission product inventory were obtained from references 2.b. and 2.c. , andI i

1 adjusted for reactor runs of approximately one hour.

b.

three postulated The radionuclides sources: inventory, listed in Tabla 2-1, was derived from (1) i.e., 3600 MW-s, Fission products produced by a one medawatt run of one hour duratio (2)

Fission products produced by a one kilowatt steady-state operation of 100 days duration; i.e. , 8640 MW-s, and (3) operations. Activation products, produced primarily from the stesd'y-state 1

All three sources are assumed to contribute to any emergency condition.

c.

It is assumed, conservatively, that self-absorption within the reactor core approximately equals backscaccer and other albedo effects. Therefore, the core is treated as approximately a point source, for any receptor located as auch as five meters from it.

d.

from reference 2.d.

Most of the special radionuclides data used in this report were obtained -

Table 2-1. Gamusa-Emittins Radionuclides Inventory I Radionuclides Curies Group Specific Produced SR Exposure Rate, 1/h @ la Ti-h t+10m (*l) t+24h (*1)

FS(*2) Sr-91 321 0.39 125 23 Y-92 8237 0.125 1030 9 Y-93 335 0.05 17 3 Zr-95 '213 0.38 81 30 Nb-97 90 0.35 32 0 Ru-105 123 0.34 42 1 Sb-129 90 (2.0)(*3) 180 6 I-132 219 1.18 256 i Enclosure 2

. _ ~

w I

Table 2-1. Gamma-Emitting Radionuclides Inventory Continued Radionuclides Curies R-m

-I Exposure Rate. R/h @ Im Group Specific Produced TtT t+10m (*1) t+24h ( *1) 3 FS(*2) I-133 177 0.26 46 21 E I-135 14 0 99 14 1 Xe-135 658 0.14 92 15

(*4) 800 0.7 560 280 FL(*5) Zr-95 32.9 0.41 13 13 Nb-95 20.9 0.42 I

' 9 9 Ru-103 25.1 O.26 7 7 I-131 25.2 0.32 6 6 Te-132 36.9 0.22 8 8 Ba-140 51.5 1.24 64 64 La-140 51.3 1.13 58 58 Ce-141 43.0 0.04 2 2

(*4) 200 0.7 140 140 AP(*6) Na-24 4.3 1.84 8 3 Mn-56 19.7 0.83 16 I

1 Fe-59 0.3 0.64 1 1 Co-60 1.8 1.30 3 3 Cu-64 13.8 0.12 2 1 I Co-58

(*4) 50 6.5 0.55 1.0 Total exposure rate: 2866 4

50 50 807 1

I (*l) t= time of shut-down (t+10 minutes, C+24 hours)

(*2) fission products from IMW square wave, for one hour

(*3) conservatively estimated from published data (reference 2.d.)

I (*4) contributions from radionuclides, not listed because of very low gamma levels and/or small amount produced.

(*5) steady-state, ikW,100 days, fission products produced.

(*6) activation products (assume 100 kg of steel in neutron field).

2. Exposure Rate Discussion.

a. All exposure rates are derived from a postulated level of 3000 R/h, at 1m. Figures 1A and 15 show the postulated positions of exposed personnel.

b. Assumed Area Par meters:

(1) reactor located 16.5 ft. below reactor deck, with maximum pool width of 12 feet.

2 (2) gamma energy = 1 MeV, so tha u/ M.0637 cm /g, buildup factors as follows (for concrete):

I #x: 1 2 4 7 10 15 20 BF: 2.0 3.5 6.8 14 23 40 61 2

(3) density (o) = 2.3 (concrete), and 1.7 (soil) g/cm3 , so that the number of mean free paths ; i.e. , the magnitude of ux, for one foot is: 4.67 (concrete) and 3.30 (soil); see reference 2.d.

c. Iccurate skyshine estimates were not obtained. Therefore, conservative estimates were used for skyshine contributions, based on reference 2.e., and with the following assumptions:

(1) energy is roughly equiva(ent to cobalt-60 (1.25 MeV),

(2) a source located at a distance of 7 feet from a receptor, located on the opposite side of a barrier which is 4 feet high, and (3) essentially all exposure results from secondary photons produced by compton scatter with air molecules ("skyshine").

The resulting skyshine is, therefore, estimated as 2% of what the direct radiation would be at 7 feet, if there were no barrier.

3. Summary of Exposure Rate Levels.

a. Exposure rates are postulated as arising either from unattenuated radiation which has traversed part of the biological shielding material of the reactor, labeled as DIRECT, or from eadiation scattered from the air located above the reactor deck, including that outside the building, labeled as SKYSHINE.

b. Exposure rate data are sununarized in Table 2-2. All locations referenced are shown in Figure 1, with the " top view points coded A, and " side view" Points coded B.

Table 2-2. Reactor Deck Area Exposure Rates Location Dist. (feet) Shielding (ux) R/h DIRECT R/h SKYSHINE A-1 23 33 =0 0.6 A-2 24 56 =0 0.2 A-3 40 99 =0 0.1 A-4 39 105 =0 <0.1 A-5 30 49 =0 0.2 A-6 <0.1 1.5 I

19 8 A-7 61 58 =0 <0.1 ,

. B-1 15 --

133 =12 {

B-2 33 =3 30 .

)

B-3 35 25 =3 l 3

' " ~ ~ ~

i i

- ._ __ _m. ,

+e  % em o I

i i

i _ I I I 7 [I l O _ . A Q' l , )

1. 1 v v n L. I n FrA ET52 UECY N A D i d TI A N L - V in L 3, byi5 gp ( -

y_

f 0h YIEW 35m I

=

Va&ri A

s na esw-r- ,A,

- t 6 ,

$ $ O f.'D f N&{ Q I >

l

• w v

. . . . 6 u.< s .

as I

B: m j 4

J L > I'pt ' } v s ==

A' 7 A 3

/ g

)

e i l i f W'

) L Ic o s

C I

i A

MO'ss j i MF AO/ _

.. 4 I~ I I

I l f

h I

l h *4uma e 0 .

I .

i I h

}

e M h M /'

i i & 3 I D fii V I & LM .

6 l i

.* A

• M-e

d' &

t I . .

i r f I J e.I. .

6 i

I i

3 (s

.6 a

i

}

1 l

y .

1 i

=#

es I

e

i. .

4 i

i i  !

l *{ e i ii n A

, g N il K114e e4X =

f i V 1 a t'il . .

a f '

rIeMC P/ A di cT '

t; ,

l 6 ., , .

sw f

7 '

I o i f

• ! ..%V

-- -- - -. - 1

. ~~-..---:- ~, ~ __.._. -_ -

__ l

4. References.

2.a. AFRRI-TRIGA Final Safeguards Report (revised March 1962), Chapter VI, " Hazards Analysis".

2.b. D. H. Slade, Meteorology and Atomic Energy: 1968, U.S.A.E.C., page 314 (Table 7.1) .

2.c. Etherington (edit. ), Radiation Hygiene Handbook, pp. 2-2 to 2-5, 7-16. ,

2.d. U.S.P.H.S. , Radiological Health Handbook, (both 1960 and 1970 editions),

numerous pages.

2.e. Price, Horton & Spinney, Radiation Shielding, Pergamon, 1957, pp.

56-60.

)

I J;

1 i

i i

s s

i i

4 I

1 I

~ -~- -- -- -

_ __ -_ _____-________r___r- _A

1 FISSION PRODUCT RELEASE: Loss-of-Coolant l

1

1. General.

a. The expected fission product inventory (FPI) af ter 100 days operation I' is given in Table 3-1, based of data from Re *erence 3.a.

by the appropriate maximum permissible conesseration in air (MPC This quantity, divided

) is the hazard index (HI).

b. The radionuclides in Table 3-1 include ggre than 99.9% of the total HI. Those radionuclides with HI less than 2x10 are not included in the final release calculations of Table 3-2.

l j .

i 2. Total Release. '

s j

a. The activity released from the fuel elements is assumed to be one percent of the total; this represents 87% of the total activity in one fuel e leme nt , 9% of the activity in 10 elements, etc. In view of the statements made in Reference 3.b., this is a conservative assumption.

b. Release from the building is assumed to have occurred, on the average, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter shutdown. The reactor building provides confinement, but not containment. It is assumed that 100% of gases, and 10% of other. fission products, will escape. Given the release conditions, and the information provided in Reference 3.c. , this is a conservative assumption. It is also consistent with the conservative assumptions required by USNRC license R-84, for the TRIGA reactor.

c. The release rate, in Ci/s, is averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, as permitted by 10CFR20.403 (see reference 3.d.).

Table 3-1, TRIGA Fission Product Summary Radionuclides  % of F.P.(*I) KPC (UCi/ol)

• ky(* } Half-life Inventerv( *'}

Sr-89 5.39 3 (-10) --

52.7d 1005 Sr-90 'O.08 3 (-11) --

27.77 14.4 Y-90 0.07 3 (-9) 0.050 64h 14.4 Y-91 6.41 1 (-9) 0.137 58.8d 1240 l

Zr-95 6.08 1 (-9) 0.366 65.5d 1175

Nb-95 3.93 3 (-9) 0.035 35.0d 1175 Ru-103 4.63 3 (-9) 0.250 39.3d 896 Rh-103 4.63 2 (-6) 0.023 57m 896 Ru-106 0.14 6 (-9) -- 368d 26.9

, Rh-106 0.14' 2 (-9)(*5) 0.503 2.2h 26.9 8 (-9)

Ag-111 0.03 0.014 7.5d 5.4 i Su-125 0.02 3 (-9) 0.165 9.4d 3.6 Sb-127 0.14 6 (-9)(*5) 0.457 3.8d 2810

. Enclosure 3 ll l .__ __

1 Table 3-1. TRICA Fission Product Suc:ca rv Continued

]

Radionuclides  % of F.P.(* MPC (uci/mi)(* } ky(* ) _ Half-life Inventory (*')

Te-127 0.17 3 (-8) 0.004 9.4d 3400 Te-129 5.01 1 (-7) 0.135 68.7m 50.4 I-131 4.46 1 (-10) 0.192 8.05d I' Xe-131 0.04 4 (-7) 0.002 11.8d 900 8.9 Te-132 6.57 4 (-9) 0.117 77.7h 1:20 I-132 6.57 3 (-9) 0.904 2.26h 1320

~

Xe-133 9.81 3 (-7) 0.016 5.27d 1970 l

Cs-137 0.06 5 (-10) 0.333 30.0y 10.7 Ba-140 9.15 '

1 (-9) 0.209 12.8d 1840 Ii I La-140 Ce-141 9.10 7.63 4 (-9) 5 (-9) 1.023 0.038 40.2h 32.5d 1830 1535 Pr-143 8.00 6 (-9) ---

13.6d 1610 I, '

Cr-144 Pr-144 1.75 1.75 2 (-10) 5 (-9)(*5) 0.014 0.018 284d 17.3m 352 352

. Nd-147 3.87 8 (-9) 0.069 11.id 779

' Pm-147 0.23 2 (-9) ----

2.62y 46.1 Eu-156 0.02 3 (-9) 0.895 15.2d 3.9

(*1) From Refergnce 3a. , supplemented by comparison data from Reference 3.b., 3.c.

I ~

(*2) (-n) =x10 ; for unrestricted areas.

(*3) ky= ganama radiation level from point source, rad-m /Ci-h.

2

(*4) Invectory at time of shutdown, curies.

(*5) Calculated by JMA; others taken from Reference 3.d.

Tabis 3-2. Pestulated Fission Product Release Radionuclides Ci(0)(*I) Ci(24)(*1) b FPI/MPC,(*2} Ci(R)(* } Ci(B)(* Q(Ci/s)(*$}

I Sr-89 1005 992 33.1(12) 9.92 0.992 1.15(-5)

! Sr-90 14.4 14.4 4.80(11) 0.144 0.015 1.74(-7)

Y-90 14.4 14.4 4.80(9) -- -- --

I '

Y-91 Zr-95 Nb-95 1240 1175 1175 1225 1163 1163 1.23(12) 1.16(12) 3.88(11) 12.25 11.63 11.63 1.23 1.16 1.16 1.42(-5) 1.34(-5) 1.34(-5)

, Ru-103 896 880 2.99(11) 8.80 0.88 1.02(-5) i Rh-103 896 880 4.40(8) -- -- --

' Ru-106 26.9 26.8 4.47(9) -- -- --

Rh-106 26.9 26.8 1.34(10) -- ---

I~ Ag-ill 5.4 4.9 6.13(8) -- --- ---

Sa-125 3.6 3.3 1.10(9)-- - ---

56-127 2810 . 2340 3.90(11) 23.40 2.34 2.66(-5)

{ Te-127 3400 580 1.93(10) 5.80 0.58 6.94(-6) 4

, Te-129 50.4 10 I-131 900 826 9.00(12) 8.26 8.26 9.56(-5)

I l 2 I

s

_ _ _ _ _ . ~ . _ . _ _ .

.l l

i Table 3-2. Pos tulated Fission Product Release Continued l

)

Radionuclides Ci(0)(*1) Ci(24)( FPI/MPC Ci(R)

• Ci(B)(* Q(Ci/s) i Xe-131 8.9 8.4 2.10(7) 0.09 0.09 1.04(-6)

Te-132 1320 1066 2.67(11) 10.66 1.07 1.27(-7)

I-132 1320 1066 3.55(11) 10.66 10.66 1. 2 3(-4 )

-Xe-133 1970 1727 5.76(9) 17.27 17.27 2.00(-4) l  ! Cs-137 10.7 10.7 2.14(10) 0.107 0.01 1.16(-7) i Ba-140 1840 1743. 1.74(12) 17.43 1.74 2.01(-5) l La-140 1830 1734 4.34(11) 17.34 1.73 2.00(-5) l Ce-141 1535 1503 }.01(11) 15.03 1.50 1.74(-5)

Pr-143 1610 1530 2.55(11) 15.30 1.53 1.77(-5)

. Ce-144 352 351 -

1.76(12) 3.51 0.35 4.05(-4)

Pr-144 352 351 1.10(12) 3.51 0.35 4. 0 5( -6 )

Nd-147 779 732 9.15(10) 7.32 0.73 8.45(-6)

I Pm-147 Eu-156 46.1 3.9 46.1 3.7 2.31(10) 1.23(9) 0.46 0.05 5.32(-7)

(*1) Number of curies present at shutdown (0) and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter shutdown (24).

(*2) Fission product inventory (FPI), in curies, divided by the appropriate MPC ; known as hazard index (HI); (n)=x10".

(*3) Curfes released from fuel elements, assuming loss of 1% of total.

(*4) Curies released from building, assuming 100% loss of gases and 10% loss of othe fission products. ~

(*5) Average release rate, over a 24-hour period; (-n)=x10 ".

3. Atmospheric Dispersion.

a. The dispersion equation used (reference 3.f., 3.h.) assumes a release I at an elevation below that of the top of the AFERI stack.

representing the effective cross-seegional area of AFRRI at the release point, The parameter A, is assumed to be approximately 200 m .

b. The equation ist X(Ci/m ) = Q (Ci/s)/wIyIzU (m /s), with: )( the downwind concentration, U the effective wind speed, Q the average release rate (see Table 3-2) and IyIs as defined belows (1) I =a + 0.5 A/W 7 7 l

(2) I,2 ,g2g+ 0.5 A/W l (3) For AFR1r conditions, 0.5A/w = 31.8.

c. Dispersion coef ficients for Type-F conditions, the worst credible 24-hour conditions, are given in Table 3-3, based on reference 3 I

3 I. 4 i

Table 3-3. Dispersion Coef ficients and Concentrations Downwind Stardard Disp. Adjusted disp. Rel./ Conc.

distance ,c_oeff., a coeff., m Q/)(; for U=1, '

x, m Q Q Q h AFRRI conditions I ,

50 100 200 2.5 4.0 7.5 1.0 2.3 4.1 6.2 6.9 9.4 5.7 6.1 7.0 110 132 206 400 14 .t 7.1 15.5 9.1 441

• Ratio of release rate to ground level concentration (m /s)

4. Dose Estimates. .

a. The maxinum permissible dose (MPD), as specified below, is the dose commitment reizted to an exposure to a concentration of 1xMPC for one day. For the total body, MPD = 1.37 millirem, for the thfroid MPD = 4.11 millirem.

b. Table 3-4 suusnarizes significant releases considered possible'in the event of loss-of-coolant, based on data prcsented in previous tables of this report.

I Table 3-4. Dose Estimates for Released Radionuclides Radionuclides Avg. Q X(Ci/m ) MPC -Equivalents Critical Released (Ci/s) at x=50m kaleased Organ I Xa-133 I-132 I-131 2.00(-4) 1.23(-4) 9.56(-5) 1.82(-6) 1.12(-6) 8.69(-7) 373 8690 6 rotal body thyroid thyroid Xr-Eb-95 2.68(-5) 2.44(~7) 122 total body Sb-127 2.66(-5) 2.42(-7) 2 total body Ba-La-140 4.01(-5) 3.65(-7) 182 total body Other F.P. 9.70(-5) 8.82(-7) 882 total body l c. Table 3-5. Summarized the unrestricted area dose possible to persons f located 50 or 400 asters of AFERI for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the release i postulated above, assuming the wind blows toward those persons for the entire 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Table 3-5. Maximus Credible Dose Summary

' critical organ mres/da @ 50m mres/da @ 400m Total body -

1403 350 Thyroid 3724 929 l 4 I

I.

5. References.

3.a. U.S.P.H.S., Radiological Health Handbook,1960 Edition.

l 3.b. AFRRI-TRIGA Final Safeguards Report, Table VI 3. I 3.c. J. H. Rust & L. E. Weaver, Nuclear Power Safety, Pergamon, 1976, pp.101-153, "Reisase of Radioactive Materials from Reactors" (K. Z. Morgan).

3.d. Title 10, CFR, Part 20, " Standards for Protection Against Radiation",

Appendit B.

3.e. IAEA Technical Report No. 152, Evaluation of Radiation Emergencies and Accidents: Selected Criteria and Data, E. J. Vallario, U.S.A.E.C., 19 7t+ .

3.f. U.S.N.R.C. Regulatory Guide 1.111, " Methods of Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", March 1976.

3.g. D. H. Slade, Meteorology and Atomic Energy: 1968, U.S.A.E.C.

3.h. HPP 2-5, " Environmental Release Evaluations", September 1978.

I I

I.

i

1 l 8 8

TRICA Emergency Rules of Thumb I

1. External Exposure Rate and Dose Commitment.

a. The purpose of this paragraph is to provide a means of estimating I internal dose commitment, based on:

(1) presence on the reactor deck following a major release of radio-activity from the TRIGA reactor, and (2) an assumed airborne radioactivity mixture typical of TRIGA operations; i.e, primarily iodines, kryptons and xenons, with medium-to-l I short half-lives. (Reference'4.a. and 4.b.).

b. Assumptions include:

(1) radioactive aerosol volume approximating hemispherical shape, with no allowance for scattered gamma radiation (slightly conservative),

effective MPC I of released fission products calculated as 3 x 10 ~0(2) (derived from reffre(nc)e 4.a. , 4.b. , 4.c. , fairly conservative) .

2 (3) af active gamma energy of 0.7 MeV, and gamma constant of 0.4 R-m /Ci-h.

c. External exposure race.

(1) integrated from infinite cloud model (reference 4.d.), over the  ;

postulated reactor deck volume. i (2) the equation,ist exposure rate (R/h) = #I; E.R. = 27 T )( (1-e-SV)f r

-8 ~7 (3) E.R. = 2Wx0.4x3x10 x0.0474 +0.0097=3.68x10 R/h

d. Dose commitment.

(1) based on a dose commitmest of 5000 millires in 2000 hours0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br />, the expected dose from 1 MPC *

(I) is 2.5x10 ~3 rem per hour of exposure to the con-caminated atmosphere. I (2) therefore, dividing the dose commitment by the exposure race yields the following:, ,

1 mR/h is equivalent to 7 res/h I .

Enclosure 4 l

l I

I.

I

e.

Conclusion:

An exposure for one hour, to a reactor deck concentration j l which results in an external exposure rate of 1 mR/h, will give a dose coc:mitmen: l of less than 10 rems.

l' l

2. CAM Readings and Concentration.

l

a. The purpose of this paragraph is to provide a means of estimating airborne concentrations, in fission product MPC-equivalents, from continuous I air monitor (CAM) readings. J

b. Assumptions:

~

(1) effective MPC,(IJ=3x10 uCi/ml (see paragraph 1).

(2) efficiency of CAM detector is 5% (1 cpm /20dpm) for mixed fission products (3) no exposure of the detector s external radiation (conservative as sumption)

(4) pump rate = 7 CFM (200,000 G 3/ minute)

(5) essentially no radioactive decay or removal from the collecting filter during the collection period.

c. Discussion: AFERI CAN's are the fixed-filter type. Therefore, the amount of activity on the collecting filter is the product of the airborne concentration, air flow race, and total time of buildup. For this reason, all count races must be interpreted in terms of total buildup time.

d. Table 4-1 summarizes the estimated airborne radioactivity levels corre-sponding to a not count rate of 40,000 cps (80% of full scale) for several buildup periods.

Table 4-1 Buildup Time and Airborne Concentration Buildup Increase in Concentration Fraction Staygime, 50 Time (min.) epa /ainute (uci/cm ) of MPC aren (min.)

~

5 8000 3.6x10 12 100 10 4000 1.8x1,g"7 6 200 20 2000 9x10 3 400

-8 1.5 40 1000 4.5x10 800

• Dose commitment (se,e paragraph 1), assuming n_o o respirator.

2 l

1

f 1

. l l

l l

3. References.

a. Title 10, CFR, Part 20, " Standards for Protection Against Radiation, Appendix B.

b. IAEA Technical Report #152, " Evaluation of Radiation Emergencies and l

Accidents: Selected Criteria and Data", E. J. Va11ario, 1974.

c. N.C.R.P. Report #55, '" Protection of the thyroid Gland in the Event j of Releases of Radiciodine", August, 1977.

d. D. H. S1s h , Meteorology and Atomic Energy; 1968, U.S.A.E.C., Chapter 7. {

l i

I I ~

I 3

I

= .. . . .

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

. - I p

f i i*I I i i f f

'  !  ! I I 8 8 '

i l i i  !  ! ' i- - =r- i

.K w! i - -

e-,

i l l - . .-i s -

I i i

, i in . .u. a.  :-- .L ==- . i.. -r 6 y  ;

r j 4 i. .:i - 7,+ .-..:- . :.2 .t. -f- _

s  : - j

,, s i i= .

.' .: ::t=L usy L -

s e 1, s L. -  :.

y~_=... ..=.----' --  ; 4- ..

, l j }t 4.- -- [ - -5 7 ==" _

== - -

r:2._ m ---H .. -i : p- ,-  !. _4

^:

I,, ;4 I t

I .}ji

} } }-

f- _-p _

g:  ;- .

p . !- l j

' I' M ik t ~4i fa'R'It- M!

r 'I

[;L i

_._L 2 i

__._ _ ;._ i M n . _ .ub k_fL.' a _.Ni i

I:

I i j h2.

T:1

. k' k '

N i i h $.'

I y A i t .. - . . . _ i. ..

.a .(  ! _,. , 1=;

.i '

! ta *q

h.  !. 1

} M

-4 -  ;~ .,

k i 3 4U F-- 7I 4 l  : i l

~

j .

l  ! I'g- 4= -- _ :. y_ . Mpj ;- i i t! .\ _.

.P F.} -

- : g i i s.== m- .

i -.

c -i i m '+

i  ;

g $

yhr -

--yw, - ::- _ i  : 4 l\

4' _ [ .. }:  :. A,[_4 -7_.i .. , Y hy i ( *._-  % 4' r

. .].

l j'j i

]l l

c_  ! A Ii 5 a . i..

, .l i

e  ; i ,

, g ._ . . .

h g f . , ii 2

,Q

-.  ; f 1 .-- . Eii. .1

~

F. I  !

3 t l l { p'. L }. j

' ' ~

\

4 r- i

'.".w' I

O

,'_ i b k. t - N  % =. -=:,= =: . L_

i

) j j- 1- -j ' i 2: q;; . ": =Mur tp=2 4=='

e . .} h A

_ ._q m =---.= =

=-!=s , m i - I

} i-' d- - -

Il55~ 2-i-?G i: E[.E-E'  ! 1 l l ,%

i  ;  : jdi=  :  :. 21'. T. .. ~.. Zj^~..- a e f (

~ ~ ' ~

D 1 [ i l { L _.; .i - - 2 kp- .

": l [ Q Q~ b

~

l l v. - .{ t i "

> ..; _-_ . _ ... . , r w 3 . _ _ _ .

'i I L i i +-

.v f '

t

'  : [

' t 4? i A_

rm . I 6F hw i l _- . - .

e #

h, I b tb . _ . . -_ ..

.  !. i i I YD ~~ ' ~ ~ ~ - ' ~ '

' i Fl } j j

(' 7 W A i  ! ' .

u

• _ D. A" . i t  ! i 9, n 85 ")  ; O

.g' " _ _ _ . . . -.

1 l . . - . .

~

1. u "

t i i _ _ _ . . _

% i D __

't i

. a '. \F

{ t f i F

.~ g_D I , . . _ .. . _. . A I l 1 6 I

i

- %yg, A

. _ . _ _ . _ . - ._ . . _ - . _ _ _ rS, QW. ,

I i-iri ./*

e.

y%(

ag, t

' ,.4.. '

I_.

t i fY i b *.

i ,

_ 'T r

' e l

• l g g e e e e e # 4 8 i i e < *  ? . . .,  % ( ( 'L b ( . ..

" *- *, S %

.4 .. T* .T t **

% < s. X ,

y - - .

• IISI 4F e. : e w.3ahll*.=Y.*M'bMY I'M O

COPY FOR YOUR DEFENSE NUCLEAR AGENCY INf0RNAll0N ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE BETH ESDA, M ARYLAND 20814 5145 13 February 1989 ,

1 TO: Reactor Modifications File FROM: Reactor Facility Director

SUBJECT:

Safety Analyses of Modifications to the Reactor Facility at the Armed Forces Radiobiology Research Institute (AFRRI)

1. In accordance with the provisions of 10 CFR 50.59 safety analyses have been performed on several modifications to the AFRRI TRIGA Mark-F research reactor. The Reactor Facility Director (RFD) has determined that in each case the modification involves no unreviewed safety questions or changes to the facility technical specifications. The Instit,ute's Reactor and Radiation Facility Safety Committee has reviewed these and concurred with the RFD's conclusions.

2. These analyses are hereto appended for the record.

Mark Moore Reactor Facility Director CF: Chief, RSDR 1

e

e' SAFETY ANALYSES OF MODIFICATIONS TO THE REACTOR FACILITY AT THE ARMED FORCES RADIOBIOLOGY i

RESEARCH INSTITUTE DECEMBER 1988 I I l 4

I I

I I

I

i l

l I I have determined that modifications to the Reactor Facility, as dese:ribed in this Technical Report (Safety Analysis Report per 10CFR50.59), involve no unreviewed safety questions and, in fact, j

are improvements to the operational capability of the facility (

I and radiological safety at AFRRI. I submit this Technical Report to the Reactor Radiation and Facility Safety Committee (RRFSC) l i

I for review and concurrence.

d A Mark Moore I Reactor Facility Director I The RRFSC has reviewed this Technical Report and concurs with the I determination that the Modifications to the Reactor Facility, as described in this Technical Report, involve no unreviewed safety questions.

N Y Richard I. Walker I CAPT, MSC, USN Chairman, RRFSC APPROVED g . .

v.

/ ~

George Irving, III Colone , USAF, BSC Director I

I I

I I .

l ABSTRACT l

This report describes changes to the reactor facility at the Armed  !

Forces Radiobiology Research Institute (AFRRI) in Bethesda, Maryland. Classified as a Safety Analysis Report (SAR) that meets the requirements of Title 10, Code of Federal Regulations, Part 50.59 (10 CFR 50.59), this document provides the basis for the conclusion that the changes to the facility involve no unreviewed I safety questions and, in fact, are improvements in the operational capability of the facility and radiological safety at AFRRI.

order to accomplish these changes, the SAR must be modified.

In The body of this report contains a complete description and detailed safety analysis of each of the SAR changes. Excerpts from the SAR l and the proposed changes are included as appendices.

Under 10CFR50.59, a licensee may make changes to its facility provided that no changes are made to the Technical Specifications, and that there are no unreviewed safety questions. The conditions for unreviewed safety questions are outlined in 10CFR50.59.a.2, and are summarized below:

If the affected equipment is related to safety:

1. The probability of occurrence or the consequences of an accident or equipment malfunction shall not be increased.

ii. The possibility for an accident or malfunction of a different type than previously evaluated in the SAR shall not exist.

iii. The margin of safety as defined in the Basis for any Technical Specification shall not be reduced.

I I

TABLE OF CONTENTS I. Introduction II. Facility Modifications Safety Analyses

1. Primary Continuous Air Monitor

2. Stack Gas Monitor System

3. Cerenkov Detector

4. Digital Voltmeter Appendix A: Listing of Corrections to be Made to the SAR Appendix B: Specific SAR Word Changes based on the analyses contained in this report.

l lI lI I

I I

l

l 1

I l

INTRODUCTION Present conditions at the Armed Portes Radiobiology Research Institute (AFRRI) require that certain modification be made to improve the reactor facility. The changes being made to the i Safety Analysis Report (SAR) include: the primary Continuous Air I Monitor, the Stack Gas Monitor System, the Cerenkov detector, and the Digital Voltmeter. j The Code of Federal Regulations (Title 10. Part 50.59) requires that modification of a portion of a licensed facility as described in the facility SAR be documented with a written safety I analysis. Such documentation p.ovices the basis for determining that the change does not involve an unreviewed safety question.

Based on the analyses in this Technical Report, it has been I determined that the pret;osed changes to the Reactor Facility do not involve unreviewed safety questions and will actually improve the operational capability of the facility and radiological safety at AFRRI.

This technical report describes changes and modifications made to the AFRRI reactor facility as depicted in the facility's SAR.

These changes have been reviewed by the Reactor Facility Director and found to contain no unreviewed safety questions. This report is submitted to the Reactor and Radiation Facility Safety I Committee (RRFSC) for their concurrence that conditions of 10CFR50.59 are met. These conditions are that no unreviewed safety questions are present and that the changes made do not incree.se the probability of occurrence or the consequences of an I accident or malfunction.

The proposed modifications require minor administrative changes in the SAR. The body of this report contains a complete description and detailed safety analysis of each of the 10 CFR 50.59 SAR changes. Appendi.x A contains a specific page/section I index of all of the SAR changes. Appendix B contains excerpts from the SAR, for each of these 10CFR50.59 modifications, and the proposed changes to the SAR.

I 3

FACILITY MODIFICATIONS SAFETY ANALYSES

1. Primary Continuous Air Monitor A flashing visual light installed on the reactor auxiliary instrumentation con >le gives illumination when the primary reactor room continuoun air monitor (CAM) is set in the TEST mode during testing. This flashing visual light is connected to the CAM through switches or relays and is designed to be activated when the CAM is in the TEST mode. Using an output signal from the Test Circuitry in the CAM to activate the flashing light does not degrade the CAM's ability co perform its intended function.

The Technical Specification Basis (3.5.1 ) for t1e radiation monitoring system is "... to characterize the 1ormal operational radiological environment of the facility and to aid to evaluating any abnormal operations or conditions. The radiation monitors provide information to the operating personnel of any e>isting or impending danger from radiation, to give sufficient time to evacuate the facility and take necessary steps to prevent the spread of radioactivity to the surroundings'.

A flashing visual light will alert the operator when the CAM is in TEST mode. The reactor is not permitted to operate with the CAM in any mode other than the OPERATE mode. The installation of this flashing visual light will improve the operational capability of the reactor facility without degrading the CAM's ability to perform its intended function. Therefore, there exists no unreviewed safety question for this item.

2. Stack Gas Monitor System A flashing visual light installed on the reactor auxiliary instrumentation console gives illumination when the Stack Gas Monitor (SGM) pump motor is turned off or the SGM ow j count warning alerts. The flashing visual light is connected to l the SGM through switches or relays and is designed to be activated under these two conditions. Using output signals from the SGM to activate the flashing light does not degrade the SGM's ability the perform its intended function.

l The Technical Specification Basis (3.5.1 ) for the j radiation monitoring system is "... to characterize the normal l operational radiological environment of the facility and to aid  !

to evaluating any abnormal operations or conditions. The radiation monitors provide information to the operating personnel

, I of any existing or impending danger from radiation, to give sufficient time to evacuate the facility and take necessary steps to prevent the spread of radioactivity to the surroundings".

l I

4

. l

A flashing visual light will alert the operator when the SGM pump motor is turned off or when the SGM low count warning l

alerts. In the event of any of these two conditions, the reactor '

is not permitted to operate. The installation of a flashing visual light will improve the operational capability of the reactor facility without degrading the SGM's ability to perform its intended function. Therefore, there exists no unreviewed safety question for this item.

3. Cerenkov Detector Sections 4.11 and 4.11.3 are being changed to include a Cerenkov detector. This change will improve the operational capability in accurately monitoring the output of peak power (NV) and integrated power (NVT) during pulse operations.

The Cerenkov detector is a photocell enclosed in an aluminum housing mounted j ust above the pool water level, directly above the reactor core. This detector was placed above the core in addition to other detectors located underwater j ust above the reactor core. The intercomparison of data between the Cerenkov detector and gamma ion chamber during pulse operations showed that the gamma ion chamber has a limited upper range due to its physical design limitations and that the Cerenkov' detector performs quite well in these upper ranges. The addition of the Cerenkov detector allows the operator to measure the power output of larger pulse operation more accurately. With the Cerenkov detector optimally positioned to measure the larger pulses, the gamma ion chamber can be adjusted to increase t.ae accuracy of measurement of power output of low pulse operations.

The Cerenkov detector and other ion chambers were evaluated in the AFRRI Final Safeguard Report dated March 1962 in conj unction with the issuance of AFRRI's first reactor license from the Atomic Energy Commission in 1962. These instruments were in operation for twenty years. In early 1983, the cerenkov detector was taken out of the reactor core because it was not being used at that time. In 1988, the Cerenkov detector was put back into the reactor core and calibrated prior to its use for pulse operation.

The Technical Specification Basis (3.2.1 ) for the channels monitoring the reactor core is "... the power level channels assure that radiations indicating reactor core parameters are adequately monitored for both steady state and pulsing modes of operations". Since the Cerenkov detector has been evaluated in AFRRI Safeguard Report dated March 1962 and was in operation for twenty years prior to its removal in early 1983, the re-installation of this detector in 1988 simply restored the reactor to its original configuration. Therefore, there exists no unreviewed safety question for this item.

5

I 4. Digital Voltmeter Section 4.11.1 of the SAR is being changed to include a digital voltmeter to give parallel readings, if necessary, to the strip chart recorder located on the reactor console. The SAR j states that "The power level is scaled on the strip chart recorder between 0 and 100 percent of the power indicated by the power range select switch on the console." The digital voltmeter I which gives a more precise output reading from the multi-range linear channel, is used to assist the operator in maintaining a particular reactor power level.

The digital voltmeter is connected into the test points on the linear amplifier circuit provided by the manufacturer for the purpose of measuring the output of the multi-range linear I channel with better precision than can be obtained from the linear pen. In addition, this digital voltmeter is used by the operator to make precise measurements at any other set of test points in the console circuitry. Since the addition of a voltmeter, which is isolated, does not alter the multi-range linear channel output, and actually gives the operator a more precise reading when monitoring the reactor power, the addition of this voltmeter actually improves the operational capability of the reactor facility. Finally, the addition of a digital voltmeter is not used as any basis for any Technical I Specification item. Therefore, there is no unreviewed safety item for this item.

I I

I I

I 6

I

~

l l

Appendix A Listing p_f, Corrections to be made to the SAR I

I ,

I

'I I

I I .

Page Section Change 3-39 3.6.2 Changed to include a flashing visual light

. I' activation when the primary CAM is in TEST mode. See 10 CFR 50.59 write-up.

3-43 3.6.3.3 Changed to include a i flashing visual light I activation when the stack gas monitoring system is in low count alert or its pump motor is turned off. See p 10 CFR50.59 write-up.

l 4-22 4.11 Changed to include a i Cerenkov detector. See 10  !

CFR 50.59 write-up.  !

4-24 4.11.3 Changed to include a cerenkov detector. See 10 CFR 50.59 write-up.

4-23 4.11.1 Changed to include a digital voltmeter. See 10 CFR 50.59 write-up.  ;

~

I 8

i I

APPENDIX B j i

specific SAR word changes based pn the Analyses '

contained .in this report.

I I

I

]

I I .

l I

9 I

1 I -

1

. k 9

1. PRIMARY CONTINUOUS AIR MONITOR Section 3.6.2 of the SAR '

CURRENT SAR WORDING:

"A description of the CAM's alarms, locations, and read-outs is given in Table 3-2 and Figures 3-12 through 3-14. The alarm set points can be found in the appropriate AFRRI internal documents.4" PROPOSED SAR WORDING:

"A description of the CAM's alarms, locations and read-out is given in Table 3-2 and Figures 3-12 through 3-14. The alarm setpoints can be found in the appropriate AFRRI internal document.4 Additionally, a flashing visual light on the reactor auxiliary instrumentation console in the reactor.

control room will be illuminated when the primar reactor room CAM is set in the TEST mode during testing.y'

2. STACK GAS MONITOR SYSTEM Section 3.6.3.3 of the SAR CURRENT SAR WORDING:

"The stack gas monitor system is capable of activating alarms at two levels."

PROPOSED SAR WORDING:

"The stack gas monitor system is capable of activating alarms I at two levels. Additionally, a flashing visual light on the reactor auxiliary instrumentation console in the reactor control room will be illuminated when the stack gas monitoring system low count warning is activated or when the stack gas monitoring system pump motor is turned off."

3. CERENKOV DETECTOR Section 4.11 of the SAR CURRENT SAR WORDING: ,

"The AFRRI-TRIGA reactor core is monitored by six detectors.

One thermocouple from each of the two instrumented fuel elements comprise two of the six detectors. A fission detector and three ion chambers comprise the remaining I 10

t reactor detectors. These six detectors are utilized to provide six independent " channels" which monitor the power level and fuel temperature of the core."

PROPOSED SAR WORDING:

"The AFERI-TRIGA reactor core is monitored by a variety of detectors. One thermocouple from each of the two instrumented fuel elements comprise two of the detectors. A fission detector, two or more ion chambers, and a Cerenkov detector comprise the remaining reactor detectors. These detectors are utilized to provide at least five independent

' channels' which monitor the power level and fuel temperature of the core during steady state operation and at least three independent ' channels' which monitor the power level and fuel temperature of the core during pulse operations."

Section 4.11.3 of the SAR CURRENT SAR WORDING:

"High flux safety channels one and two report the reactor power level as measured by three ion chambers placed above the core in the neutron field.

High flux safety channels one and two are independent of one another but operate in identical manners during steady state operation. Each channel consists of an ion chamber placed above the core and the associated electronic circuitry. The steady state power level, as measured by the two high flux safety channels, is displayed on two separate meters located on the reactor console.

During pulse operation, high flux safety channel one is shunted and the sensor for high flux safety channel two is switched to a third, independent ion chamber placed above the I core. High flux safety channel two measures the peak power level achieved during the pulse (NV channel)."

, PROPOSED SAR WORDING:

"High flux safety channels one and two report the reactor power level as measured by independent power monitors (ion chambers or Cerenkov detector) placed above the core.

High flux safety channels one and two are independent of one another but operate in identical manners during steady state operations. Each channel consists of an ion chamber placed above the core and the associated electronic circuitry. The steedy state power level, as measured by the two high flux I safety channels, is displayed on two separate meters located on the reactor console.

11 i

i

f. #

l L ..

During pulse operation, high flux safety channel one is

shunted and the' sensor for high fluxLsafety channel two is n switched to a third, independent detector. High flux safety channel two measures the peak power level achieved during the pulse (NV channel)." ;l

\-

l

l. 4. ' DIGITAL VOLTMETER Section 4.11.1 of the SAR CURRENT SAR WORDING: i "The-power level is scaled on the strip chart recorder between 0 and 100 percent of the power indicated by the power range select switch on the console. The strip chart records this output for all steady state modes of operation but not during pulse operation."

. PROPOSED SAR WORDING:  !

"The power level is scaled on the strip chart recorder or indicated on'a digital voltmeter for precise reading between 0 and 100 percent of the power indicated by the power range select switch on the console. The digital voltmeter is connected into the test points on the linear amplifier circuit provided by the manufacturer for the purpose of i measuring the output of the multi-range linear channel. The j strip chart records the output for all steady state modes of operation but not during pulse operation."

'l i

i l 12 l

l

i 1

1 k

l

[-

1 l ATTACIIMENT D LICENSE EVENT REPORTS 1

I 1

u ,

l.

l 5

fg DEFENSE NUCLEAR AGENCY 4 ARMED FORCES RAoloBloLoGY RESE ARCH INSTITUTE SE TH ESo A, MARY LAND 20814-5415 Memorandum 23 March 1988 l LICENSEE ' EVENT REPORT (LER)

Abstract: The Reactor room primary Continuous Air Monitor (CAM) was accidently left in TEST mode during the reactor operations on 29 February 1988. Two sets of corrective action were teken to prevent reoccurance. One was to instruct the Safety and Health Department staff to be more attentive to their routine maintenance on the Reactor CAMS. The second was the installation of a flashing visual light on the reactor auxiliary instrumentation console in the control room. This flashing light is illuminated by a signal from the Primary Reactor Room CAM when the TEST circuit is activated. This will alert the operator when the CAM is in TEST mode.

Circumstances surrounding the event - 29 February 1988:

0640 Startup checklist completed, all required systems are operable.

0834 Console locked by reactor operator after two experimental runs; reactor operator proceeded downstairs for an Exposure Room opening.

0845 Routine maintenance f reactor CAM's in-progress by a member of the Safety and Health L partment (SHD). During this maintenance, the Reactor Room Primary CAM was placed into TEST mode by a member of SHD, a non-licensed individual, and was accidently left in TEST mode upon completion of the routine maintenance (SHD routinely performs the maintenance of all CAM's in the Reactor Facility).

0921 Reactor operator returned to the control room to continue with the experimental runs; the reactor operator did not know that the SHD member had performed a daily check on the CAM, and that the CAM had been accidently left in TEST mode.

The reactor operations for the day, after 0921, included a series of one minute runs at low power of 1 Kw (five runs total), and a series of medium power pulses between 82.05 and $2.15 step reactivity insertions. During the pulse mode operations in the afternoon, the reactor operator on console first noticed e.bove normal CAM readings on the readout meter in the control room, but attributed these above normal readings to those levels which are expected for medium power pulse operations.

At 1521, while doing the Shutdown checklist, the reactor operator printed out an hourly report from the Stack Gas Monitor, which is adjacent to the CAM on the Reactor deck. The Reactor Operator then noticed that the tracing on the strip chart on the CAM was not commensurate with the operations conducted i

Memorandum 23 March 1988 Page 2 during the day; the CAM trace on the CAM chart recorder was a straight line trace as would be expected from a test input signal, instead of a fluctuating trace as would be expected from a series of medium power pulse operations. Upon further inspection, it was discovered that the CAM was in TEST mode and had y been accidently left in TEST mode by the SHD member upon completion of routine maintenance that morning.

This event was due to a Safety and Health Department staff member not following the proper procedures to turn the CAM back to the OPERATE mode after completion of routine maintenance on the CAM. In addition, due to the type of operations performed that day, the reactor control room CAM Readout Monitor levels appeared approximately correct when observed by the Reactor Operator.

This was particularly significant as the Startup Checklist was completed earlier that morning, ensuring that all required instruments were operable.

The event was reported to the Reactor Facility Director, who notined the USNRC telephonically, the following morning.

Probable consequences:

Leaving the CAM in the TEST mode during reactor operations would not enable the reactor ventilation system to be automatically secured via closure dampers by a signal from the CAM if the high alarm setpoint had been reached. However, manual closure of the dampers would occur if initiated by operator action. The reactor operations during the day of the incident consisted of a series of short, low power experimental runs and a series of medium power pulses, and thus any release of airborne radioactivity would have been unlikely. In addition, none of the Reactor Room Radiation Area Monitors (R1, R2, R3, RS) nor the Reactor Stack Gas Monitor registered any above normal readings, consistent with the reactor operatic ns throughout the day. The Reactor Stack roughing and HEPA filter systems wetre functional throughout the day and were capable of performing mitigation had a release occurred. Based on the above, there were no radiation releases due to the incident and no adverse effects on the facility.

Status of corrective action - 1 March 1988:

The event was reported by the Reactor Facility Director to the USNRC, Region I  ;

(Curtis Cowgill) by telephone at 1130.

Two sets of corrective actions were taken to prevent recurrence. The first was to instruct the SHD staff to be more attentive to their routine maintenance on the Reactor CAMS. The second was the installation of a flashing visual light on the reactor auxiliary instrumentation console in the control room. This flashing light is illuminated by a signal from the primary Reactor Room CAM when the TEST circuit is activated. This alerts the reactor operator when the CAM is in TEST mode.

i c.

Memorandum 23 March 1988 Page 3

Reference:

There have been no previous similar events at this Facility.

Point of

Contact:

Reactor Facility Director, M. L. Moore (202) 295-1290. s f.

)

George rving, III Colone , AF, BSC Directo d

4

' DEFENSE NUCLEAR AGENCY ARMED FORCES RADIOBIOLOGY RESEARCH INSTITUTE i l

BETHESDA MARYLAND 20814 5145 DIR 9 November 1988 Nuclear Regulatory Commission ,

j Region I )

Office of Inspection and Enforcement

)

475 Allendale Road

.)

1 King of Prussia, PA 19406 l

SUBJECT:

Licensee Event Report Gentlemen:

In accordance with 10 CFR 50.73, the attached Licensee Event Report is submitted for your consideration.

The point of contact for further information concerning this event is the Reactor Facility Director, M. L. Moore, (202) 295-1290.

GEORG . IRVING, I Colo , USAF, BSC Direc or CF: Herb Williams I

DEFENSE NUCLEAR AGENCY ARMED FORCES RADIO 810 LOGY RESEARCH INSTITUTE BETHESDA, MARYLAND 20814 5145 RSD 8 NOV 88 LICENSE EVENT REPORT (LER) f Abstract: On Oct 11, 1988, the reactor key was placed in the "on" position to review a facility circuit. A short while later the senior reactor operator who had unlocked the console while tracing an electrical circuit left the control room to retrieve an electrical schematic located in an office directly across the hall. Another reactor operator entered the control room, noticed the console was unlocked, locked the console and removed the key.

Two corrective actions were taken in response to this event to prevent reoccurrence. One, the reactor operator was counseled by the Reactor Operations Supervisor (ROS) and the Reactor Facility Director (RFD) on his responsibilities as a senior operator. Two, all the reactor console circuit schematics will be available in the control room.

Circumstances surrounding the event - 11 October 1988:

1335 Reactor operator unlocked the console to perform electrical circuit tracing and testing. Schematics used to identify circuitry in the console were obtained from a book shelf in the control room. While tracing a circuit the operator discovered that the other necessary i schematics were in a different room across the hall.

1344 The reactor operator left the control room to retrieve the schematics. i 1345 Another reactor operator entered the control room, removed the key from the console, and notified the ROS.

The reactor operator who unlocked the console was out of the control room for less than one minute before another reactor operator locked the console. The reactor was shutdown, and there were no operations, core maintenance, or fuel handling, in progress. The only personnel in the reactor were four other reactor operators and one trainee.  ;

The event was reported to the Reactor Operators Supervisor, who then notified the RFD and the USNRC that afternoon, on the same day.

l

LICENSE EVENT REPORT (LER) 8 NOV 88 Page 2 Probable Consequences:

With the reactor in a shutdown condition and no fuel handling, core maintenance, or operations being in progress, there was no reactivity change of the reactor core during this period of time.

The reactor power chart recorder indicated no power increases above normal source level during the time when the console was

)

unlocked. This information verifies that there were no consequences to the reactor facility or personnel during this time.

1 Status of corrective action - 12 October 88 Two corrective actions were taken to prevent a reoccurrence.

One, the reactor operator was counseled by the Reactor Operations Supervisor (ROS) and the Reactor Facility Director (RFD) about full compliance with regulations and his responsibilities as a i senior operator. Two, all the reactor console circuitry schematics will be available in the control room.

Reference:

There.have been no previous similar events at this facility.

7sint of

Contact:

Reactor Facility Director, M. L. Moore (202)

'95-1290.

I Mark Moore Chairman Radiation Sources Department Reactor Facility Director

I i'

. lI LI .

I ATTACHMENT E v RRFSC APPROVAL I SPECIAL REACTOR AUTHORIZATION OF 6

I I

I I

.I I

~

I I-

-I I

I I __m ____ _ ___ _ - . - _ _ _ _ - - _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - ____m_______..-__m____.____________ 2____________.

1 D,.,ISP.OS,.,ITION.-e. -u FO ,,RM

... o . . - i ... + . 1.co.

AETE;.EP.CE On Orrir4 STWOOL $U6 JECT RSD Special Reactor Authorization # 20 70 FROW DATE Cutt Files maiman, RSD 20 October 1988 {

(RFD) FELTY /jrf/51290 l

1. On 19 October 1968 the Reactor an'. Radiation Feility Safety Ccs:xnittee 1 (RRFSC) approved the use of M* tyye IR115 film for irradiation in reactor I exposure feilities for radiographic p.trpo,es. e The attached documentation contains the package that was reviewed by the comittee along with the j comittee's approval voting sheet. l l

2. This Special Reactor Authorization has been assigned: Special Reactor I Authorization # 20.

N MARK L MXRE i

chairman, RsD Reactor Fei1ity Director

{

f l

l i

l l

1 I i

i

' ' " ' ' ~ ~ ' " " '

DA 2% 2496 i

DISP,w.SITION r., .

FORMO mue-is:a.ai

v. i.nr acy raco.

rrence.cc on omcc smooi. sus 4cr f RSD Request for SPial Reactor Authorization i

vo rnou care curt Reactor & Radiation Facility Mark Moore 19 October 1988 Safety Cannittee Chairman, RSD TDG/wwt/51290

! 1. Request a special reactor authorization be issued to place 1.6 grams thin-layer cellulose nitrate film products (Wah type - IR115 film) in reactor amwre facilities for the' purpose of arpnaing to neutron flux danaities of.10-15 x 10E2P12 neutrons per square centimeter.

The materials to be irradiated include thin layers of rat tissues with Boron 10 loading on the film. The present user includes an investigator fran the National Institute of Health.

2. Tbeze .ms zooently been a greater zesearch inte1W in using these film products which convert each dmpact by imiming particle to a separate physical 1@ which ocuM result in a hole in the sensitive film layer. 'Jhis particular type of film is inaanaitive to the effect of gamma or beta radiations that are present in neutron haams as well as to light. Therefore, neutron radiography or alpha-partiola detection can be carried out in the presence of a high beta or gamma field. 'Ihis film is being used to study baron neutron capture therapy for brain through a radiation therapy uniality, employing the administration of 10 c- - ;-3w 1, which amnailates in the tumor tissue, followed by irradiation with ic; e.-sy neutrons. Deta11M information on the efficacy of borcoated c--;-3wi= for neutron capture therapy is described in the attached article from the Brookhaven National laboratory.

3. Rimilar experiment was aypvvui ly the RRPSC in April 1979, in irradiation of old painti.ngs, for the Smithamian via the National Bureau of Standards for radiographic verification of age and authenticity. In this new film experiment, these will be no reactivity changes amaMated with the irradiation of the film and no radiological safety hazazti to the personnel.

I The only safety mnaidavation is the spontaneous ignition of the film at 180 degzees C or greater. At that temperature, over a period of time, the film would Anour a slow burning. It is not expected that the 180 degrees C limit be zuww*mi at all in this experdment.

4. Radiological safety rvmaidavations will be han41ari through the Safety and Health Department.

hfA Met xxas Chairman, RSD l

Attached: as statei c

" ' ~ ~ ~'

DA 25% 2496--_-_ _ _ - - _

i l.

I-JCIION ITEM FOR RRFSC CCH4ITTEE MEETItG 19 OCIOBER 1988 Page 2 l

Request this special reactor authorization be approved and once it is successfully performed it will be ccnverted to routine reactor authorization. ,

-x -

Reactor Facility Director s a voting sunmary of RRFSC membepg:

MG d[

DISAPPROVED ff.

' ' - - R. Walker, CAPT, MSC, USN Chaicnart DISAPPROVED A d m T. O'Bri ,.Act'ing &

[

l APPROVED DISAPPIC/ED /

M. Moom , Reactor %cility i Director, AFRRI l DISAPPROVED Dr. Marcus H Voth Pennsylvania State University APPROVED DISAPPROVED .

t t

J. N. Stone Naval Research Lab.

APPROVED DISAPPROVED J. Misner, O. R. I DISAPPROVED G'. Zemah, CE, MSC, USN.

E, M APPROVED DISAPPROVED D. Alberth, MAJ, USN ,

U.S.U.H.S. I

! I

- __ _- - _ -__--- _ i

' 1

.)

ACTION. ITEM POR RRFSC CDMITIEE MEETING'19 OCIOBER 1988 Page 3 Rarm =and Approval.

? U RICHARD I. WALKER, CAPT 01 airman, RRFSC y-l

~..

Approved.

. IRVING l ,-

Col , USAF, Dir r, AFRRI i

l

_ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ . - _ . . _ .l

l

/<

[y9

• Microanalytical techniques for boron analysis using the MB(n,a)7Li  ;

reaction *W e f Ralph G. Fairchild Maicalhpartment Brookhawn NationalLabontory, Upton, New York 11973 [

Detlef Gabel *8 l (

l M&icalhpartment, Bmokhawn NatwnalLabomtory, Upton, New York 11973 and Chemistry Apartment, UniwruryofBrenser Bremen FedemiRepublicofGermany Brenda H. Las^er arxi Dennis Greenberg

} Medical Apartn.ent. Brookhawn NationalLabomsory, Upton, New York 11973 Walter Kiszenick s P61ytechnic Institute ofBmoklyn, Brooklyn New York 11201 Peggy L Micca Medical &pertment Bmokhawn NationalLabontory, Upton, New York 11973 (Received 4 January 1985; accepted for publication 9 October 1985) 5 l

In order to predict the efficacy of boronated compounds for neutron capture therapy (NCT), it is a

mandatory that the boron concentration in tissues be known. Various techniques for o measurement of trace amounts of boron (1-100 pom) are available, including chenucal and p

physical procedures. Experience has shown that, wiu the polyhedral boranes and carboranes in is particular, the usual calorimetric and spark emission spectroscopic methods are not reliable.

7 Although these compounds may be traced with additional radiolabels, direct physical detection of boron by nondestructive methods is clearly preferable. Boron analysis via detection of the prompt-y ray from the *B(n.a)'Li reaction has been shown to be a reliable technique. Two prompt y facilities developed at Brockhaven National laboratory are described. One, at the 60 MW high flux beam reactor, uses sophisticated beam extraction techniques to enhance thermal neutron intensity and reduce fast neutron and y contammation. "Ibe other was constructed at l

Brookhaven's 5 MW medical research reactor and uses conventional shielding and electronics to

]

provide an "on-line" boron analysis facility adjacent to beams designed for NCT, thus satisfymg j one of the requisites for climcal application of this procedure. Techmeal restrictions attendant j upon the synthesis and testing of boroutad biomolecules oAen require the measurement of trace amounts of boron in extremely small (mg) samples. A track etching technique capable of detecting ng amounts of boron in mg liquid or cell samples is described. Thus it is possible to pr j to measure the boron content in small amounts (mg samples) of antibodies, or boron uptake in cells be grown in tissue culture.

gt

1. INTRODUCTION this inconsistency may be due to incomplete breakdown La g

the cages.' Spark emission spectroscopic techniques Technical difficulues associated etn the synthesis of boron- equally unreliable. Conventional radiolabels have been ated biomolecules make it necessary to measure the boron as tracers in order to circumvent problems encountered I content of these compounds at vanous stages in their pro- boron nueroanalysis; however, a further synthenc step ts duction. In addition, the many vanables involved in evaluat- j quired to incorporate the radiolabel and in addition, the  ;

ing biological effects from the "B(n.a)'Li reaction make it tential lability of the ta6 as well as of boron itself introd

! necessary to have accurate information about boron content unwarranted complications.

in tinues dunng the time course of experimentation. Per. Boron analysis via detection of the prompt y ray frocn haps most important of all, there is a Arm consensus among g ,

• B(n a)'Li rescuon has been reported by others to those who have been or are involved in clinical application of l U.S.8# and has been developed in Japan for invesogacon 1 neutron capture therapy (NCT), that knowledge of boron . NCT.2

• The 478-kev prompt y ray is emitted from the as concentration in tissue immahntely prior to irradiation of cited state of 'Li in 93.5% of the decays. This coethod sa humam is mandatory.'d quantiacation is inet=~wt of the chemical form of he It is the common experience of many of those involved in Another major advantage of this procedure is that it is er the developinent of boron compounds for NCT that conven- destructive; no observable damage is produced followmg

tional microanalytical techmques for boron analyses are un. relatively low exposures to thermal neutron beams. E satisfactory 8 Chemical colortmetne procedures require the quently, for example, boronated monoclonal annbodies conversion of boron from the stardng form to boric acid, and be evaluated for boron content, and then used for are time consuming, taking many hours to days.'In particu- expenments. Unfortunately, facilities for prompc r-

, lar, we have f(,und that the latter methods do not produce analysis oAen have low sensitivities for g amounts o(

consistent results with the polyhedral botane and carborane in tissue samples. In addition, the use of sophisocated cages commonly incorporated in expenmental compounds, , extracuan devices,Nuch as neutron guide tubes to ez 50 tees. Phys.13 (1),Jen/Pe teos

oce4 34es/es/otooewsot se @ 1ees Am, Assee. Phys. leed.

A__ -- --

)'

I 1f rweruse et s/. - Wicroaneiynas 'or ooron oy to ' sus.a , ' escuen ,

I' thermal neutrons, as well as specialized electrorucs, ewes' o a ;e emusi :s hs anticoincidence devices for background suppression,

)

gt formidable obstacles to investigators with limited EvAcuarto ruas 2

p 8-Twoprompt yfacilitieshavebeendeveloped ,,noow g ggt: one of these is at the 60 MW high Guz beam reactor \ SMPt.E Hot. DER SMPL.E BE M j

.ggFBR), and uses sophisticated (and therefore expensive) \ CArcHER!

i g,sas extraction devices to enhance neutron intensity while ,,,

annunmos fast neutron and y contammation. The second '

) g,ali.y was constructed at the 5 MW medical research reac- [onIw carscroR NEUTRON wiNoow/

' g (MRR), with conventional (and inexpensive) shielding '"* ,

. s,d electronics. With this apparatus it is possible to obtain j raped "on line" boron analyses immediately prior to possible

}

,,nent irradiations (for NCT) at the therapy beam ports of l

, se MRR. Both facilities have similar sensitivities of ~200

, wants/ min per g B, and can detect at least 1 pg B per M l % d typical *cmM 8 usaHor prunp(1 analms cPB at a psal ussue in a few minutes. pure thermal neutron beam fran BNL's high-8us beam reactor.

While prompt y measurements are adequate for boron saalysis of g amounts of human and animal tissues, they are not sensiuve enough to detect ng amounts in I (mg) sam-the H 1 beam at the HFBR is being used. The thermal neu-pisa Biochemical techniques used in the synthesis and test. tron Gux density is ~2X 10' n/cm 2s at 60 MW; sensitivity as of boronated biomolecules typically produce extumely is ~ 700 counts per g B per 200 s, which is ~ 8% of back-anal! quantities of natural boron for measurement; t rono,

' ground over the Doppler broadened prompt y peak at 478 donalanubodies are difBcult to obtain and expensive es en in kev. Thus, for the usual 200e measurement, the ermr mg amounts; various column chromatographic ana yses caused by background ( l standard deviation ) c=0.15 pg B.  !

produce multiple 'anall samples; and typical cell culturt ex. All measuretnents are obemad from 1 g samples of tissue I

perunents prod 0cc ~10' cells (~1 mg) per cell cultare and/or water in quartz test tubes. Typical spectra obtained sank. Such aliquots require boron analysis with a sensitiv9y from water and boron standards with a conventional multi-6 the order of 1 ng boron per sample. Over a period of ti ,e channel analyser are shown in Fig. 2. Background for each track etching techniques have been developed to satisfy o och run is obtained by ==mmg counts in an equivalent number squarements. Fleischer found that cellulose esters could, de.

of chana*Is above and below the prompt y peak. Counts iset charged particles selectively," thus makmg it possible to from the 478-kev prompt y ray (with background subtract-detect alpha particles from the B(n.a)7Li reaction in the ed) are then corrected for possible fluctuations in reactor presence of y rays and thermal neutrons. Lelental used this 2 power, counting time, etc., by normalizing to counts from technique to measure og amounts of boron per em that had the 2.23 MeV capture y ray from hydrogen, as indicated by been vacuum deposited on tape,ts while Thellier and colles.

the equation gues evaluated Li distribution in the brain." More recently, Larsson has used similar techniques to evaluate boron distri-

1. " # buoon in tissue sections and Gabel et al. have reported a method for evaluating ng amounts of boron in 0.5 1 drop- '

aBn psic'u pt.r kacit.ity N

yg lets." The latter procedure has been mod: Sed as detailed here to evaluate similar amounts ofboron in lysed cells. Thus a N-6 BE AM; Eo MW ; 2co sec N '" 9 3- o so **v --

the results of cell culture experiments can be evaluated on d l P

cuces >

the bas:s of known cellular boron content.

b g

o 2o *0 '08  ;

E us n.v i ll.idETHOD AND RESULTS " "

ethe '

a the h8 8 [2 if

- q At Brookhaven's HFBR, vanous beam extraction tech- ' "' i o' sn of n ex. naques, such as Ni plated glass wave gmdes and single crys- l  !

E$of tal Bi "Bruckhouse" alters, are used to produce " pencil" lgi -

1 Bron. beams of pure thermal neutron ( ~2 cm in diameter) that mon. are free of significant contammations of fast neutrons and y e up. T' rays. A stylized conaguration of such a facility is shown in g No Othe Ense. Fig.1. In this geometry, background rediation was low; 8 go,7, g,ygggf . - i -

i may shseldmg around the 2 X 2 in. solid state detector [ pure Ge o 2eo ieo ieo 2oo gc ) or Ge(Li)] was rnmtmal, and the detector could be posi, cH ANNEL. Nuwer e

y. ray noced within a few inches of the beam. Background was reduced by absorbing thermal neutrons scattered of a g, son beam sample with an enriched LisCO3 cylinder positioned sempiss, sad the response tren '*B standards a H,0 used for cabbranos d

tract manally with the beam and around the sample. Currently unknown amouse at baron is vamos insmas.

J b_ -- .

52 av: nc or u wcmacme w w:c n no w L. mm lN B, = 1 B - L - C)H., / H . (1) the 'BM.a fLi reacnon were detected and 2napec u 3

where B, = counts from 478 kev Doppler broadened peak scnbed above. At both the MRR and HFBR,2 x 2 in. cy f corrected for background and fluctuations in reactor power, drical Ge crystals were used so that relauve sensitivities w samp!c size, and positioning in beam: B = uncorrected determined prtmanly by the solid engle subtended by coun's in boron " window"; L = lower window (1/2 the detector (i.e., by the source-detector distance as given in

} width of B), just below the boron peak; U = upper window text). Specified efficiencies were - 17% (relative to 3 X 3 I ( 1/2 the width of B) just above the boron peak; # = counts Nal crystals). Resolution of such detectors can be quite from the hydrogen prompt y reaction (2.23 MeV); and ( < 2-kev full width at half maximum). However, H., = average of the H prompt-y reaction for the day's sam. used here were selected for economy rather than resolu

)

pies. as the Doppler broadened 478-kev peak precludes expl ,

Absolute calibration for boron was accomplished by con- tion of the highest resolution available. Dus actual i structing standard calibration curves using samples of boric tion was a few kev (* Co). At a power of 2 MW, the I acid (H 3BO )3obtained from the National Bureau of Stan- neutron flux density is ~ 2.7 X 10' n/cm s, 3 and the sens ,

dards.' Five samples varying in B contenis are used to con. ity is ~500 counts per 200 s (16-cm source-detector struct the calibration curve, and the true boron content ( B r ) tance), which is ~4% of background (y contamination i is obtained from the beam :::500 mR min-' or 1.29 X 10-' C kg-' min-'

Thus at the MRR, background is twice that at the HFB B, = A B r + C . (2) and the sensitivity is somewhat less. It was found that F signal-to noise ratio was improved considerably by red

It is assumed that the isotopic concentration of B in natural the angle formed by the tacident beam, sample, and de boron is 20% (' B concentration = 18.5% by weight).

5 esa esuma y t a was a conseq Access to the H 1 beam of the HF~BR is limited because of ,

of a reduction in energy of the Compton scattered ph requirements of physics expenments. As noted above, prob- 8 from the sample. Spectra from water and boron stan lems encountered in chrucal application of NCT have dem-

• measured at the MRR are shown in Fig. 4. l onstrated the necessity of deternurung boron content of b blood, and of tumor also, if possible, immediately prior to the i l irradiation of patients. Similar measurements during the B. Boron analysis by track etching ,,

course of irradiation would, of course, also be desirable. Prompt-y measurements are ideal for 1-g tissue sam 82 Consequently, an on-line facility for prompt-y analysis of However, the sensitivity of the method is not adequate for P t

boron has been constructed at the MRR radial tube ( Fig. 3 ), a mg samplea, such ." those obtamed from various ch d which is a few meters frotn the patient treatment facility. In assays as well as tissue culture experiments. Therefore a this apparatus, beam extraction, collimation, and shielding technique developed by Gabel etal.'8 has been used for I' have been accomplished by using conventional (low cost) smaller samples. In this procedure,0.5 pl drops (0.5 mg) I materials available at any reactor facility. Prompt-y's from the solution to be analyzed are deposited on cellulose ni n O

(r MRR RADIAL TUSC 8.

PaoMPT - F FACIL if Y g

REACTOR Fact GM APMif s PLUG ,

e etAM [ /a a e </Ve a < \ , " N <'

carenta "N '. ' U co n coLu. aron <4 '

j*'

}*

j l 3

{ gedAro. /g isle anscn , ,

7 , "- -

x, i

,_ y q\ <If /\s ** - s , '?*

aux , <,', ' . . $,' as.. ' c ,x 1

E me ,_ _

t, cong go, 3, g g,7 p j

'\

i ',h.' [ .' re c  ! * *8 b " #

l

~

lA tra 2

t

• C0 eve

/

\au-&&&[//$

at sm--$ g , *A/

i

, os 12

• e 0. @ LCA0 u ncos coLuuaron M somose sittL g g

'q -. 3 s -.-- wg 38--~***M' UncoptpoiY -

l conal sMerTs ,

seedoel Pityedca, Vol.13, No.1, Jen/ Pet 1998 )

mM m . . _ _ _ _ _ . _. ~ l

y n,re ao et s wero4rwn 'or seren w me 'sei a , L. , .cuan Boron is mued with F.10 cell mecum plus few wne w.

k '" 5 - rum ( FBS), wiuch serves as a matru to hold the ooron w tu;e c!

y N6 drying. Relatively umform Selds of alpha tracks are ob-

'; 478 kev SilkeV , tained in this manaer (Fig. 6). Asymmetric deposition of g4

- boron is found if water is used as a solvent; the small crystals

,in % 3 . M 'O that were observed when phosphate bufered saline M3in G higi, i> . (PBS) + 1 mg protein /ml was used as the solvent'5 were "l* 'li" "***d' Wh $' '

'H 2O AAer irradiation, the aim is developed in 10% NaOH at etion goif f: gp agg 60 *C for 45 min. Tracks are then counted optoelectronicauy esoy, ] , fotod - with a leitz microscope (4xobjective with a green alter) g  ! MRR PROMPT y FACILITY interfaced with a Quantimet image analyzer. Fields obauned g g Ra01 AL BEAM 2 MW. 200 sec from drops with I and 10yg natural boron per ml are shown I ' i i i i i i i in Fig. 6 along with background delds. As can be seen from Dr dis.

• g l#

h #80 185 190 CHANNEL 195 200NUM8ER 205 210 215 ( the calibration curve (Fig. 5), the sensitivity extends below g-i) 0.1 g natural boron /ml (0.02 yg 'D/ml). The technique 8

can thus detect ng amounts of natural boron in the 0.5-mg

!FBR y 3 4 s,.stra obtained at ila MRR. shomog backsround from wier drops used.

,,,pm and ibe rapoem from s standards in weier. Cells are analyzed for boron content following the proce-ggor, dures described for hamster V-79 cells below. It is possible to work with a " slurry" consisting of a manmum of ~ 5 x 10' een.e ask detectors (Kodak Paths LR115 type 1; 6- m thick Eton, i de on polye.ter sheets), and irradiated to -6x 10 n/

88 ceus (-0.05 mg) in 0.5 ing p20, so that the lower Hanit for boron determmation in cells is ~ l g B/g. Due to the resid-dards ,~ ,i (1s MW,300 s), at the patient port of the MRR. This ual mass (dried cells in the samples analyzed, correction for

,, diaten facility is particularly advantageous as a series of self abewpuon d the a panicle must be made. In wder to abers between the core and point ofirredsstion selectively obtain "sW" curm agamn w&h *h" i eH amoves fast neutrons and y rays in an esort to optimize the

' SamPl es could be compared, known amounts o(boren were thermal neutron dux density. For track etching techniques, I intaed in wis ceu slunys. Bwon standard solutions we

[ ' pass recoils from the interacuon of fast neutrons with hy-dresse can cause background interference. The ratio of ther.

Pmpand ming unennchah scid, water, and 10% faal bonne senun and dDuted m snake concentrations d 0.01,  ;

- nel neutrons to fast neutrons in the above facility is 50. ,

0.1,1.0, and 10.0 g 8/ml, with a control of 0.0 g B/ml. V-Here, fast neutrons are de6ned as those with energies abo e f"' 10 kev. Gamma and fast neutron done rates are ~ 15 and 25 ' D **U' ** YP",a rEmin (0.15 and 0.25 Oy/ min), respectively, at the point harmted, and counted. Ceu peb lasengd10 censwenobeM EachceupeUnwas

' afirradiation, and do not contribute sigmacant background frozen fa one 4 to anow lymng. Then,100 pl d the buon standard was added 2 mch ceu peHa. AAetwonex- l (ese lief.16 for a detailed discussion of the dosimetry).

Scandard solutions of natural boron are deposited on the pg, the aandard ceu soh waN in a ~ 2M capac-tie to construct a calibration curve, as shown in Fig. 5.

sty Hamilton syringe. Using a 0.5 91 repeat r=;- -- ~, dro-plets were placed on cellulose nitrate ilm, dried, and irra-

, disted at 1 MW for 60 s (l.2x 10'8 n/cm 8); 0.5 pl droplets

! io*, , , , , , , , , , , ,,u of the standard solution (no cells) were also added to Alm,  !

which was etched as described above.

. f 1 The calibration curve with ceus is shown in Fig. 7 (bottom q p cAUBRA oN ME -

. curve) and compared to a calibration curve obtaAned from p

I .the same 61m, without ceus (upper curve). It is apparent

. '0Y '. cELwLoSE NITRATE MLW  : that self-absorption in the cells reduces sensitivity by

' ~40%. The V 79 cells have a diameter of 12.5 pm," or a

{ ,, .

7 volume of ~ 1000 m .8 Thus 10 cells will have a volume of

,  ;,oc 10 1, so that when the 10' cells are brought up to a anal

=-

SPPan ,

• 'd* ,

!Ir ~

volume of 100 1 (water + 10% FBS) there will be 5 x 10' saio'so f .: -

cells per 0.5 I drop. When dried, the residual cellular mass

{a'3 0 ;-  : will be ~0.005 mg spread over ~ 1.3 mm8 (assummg dried i i  :  : weight:: IC % of wet weight) or ~0.33 mg/cm 2. This corre-sponds to 1.8 of unit density matenal, whkh is a sizable t .

' , , , , , , , . , , , , , , , , , , , , , , , fraction of the 10.p range in tissue of the a particle from the

,o j

oooi o.oi o.i i .o io loo B(n.a)7Li rescuon, thus producing the self-absorpuon saa a per mi . demonstrated in Fig. 7.

Rotn the abon data n Wthat mg mts M

PE 5 Cahbranos curn for ceGulose natrese slm obtamed mth 0.53 al auss of mandard solunees a(natural boros and irredmind mth 6 x 10 n/ . can be evaluated for boron content with a sensitivity down to em'. a,eksround noi suntracted. ~l g B/g of cells, and that corrections must be made for

{ Phyeles, vel 13, Ms.1.Jan/ Pet 1988

l

$4 FaercMd et s/. . Wicroana#ys4s 'or Doron my the 'Sm a & section .

.N

.t .

8 . . 1

. i I ,

a

~

L t.

(

i .

' Flo. 6. CeDulces astrase Rim irradiased 4 ' 3 s

-6 x 10'8a/cm s at the h0Ut. (a) Be around; (b) backsround from cell gror

. ..* ~

y j .m ' A , media used for suspenson of boron n- i

. , . *'J_' f dards; (c) i ps B/ml in 0.5-41 drov. (d) '

' i ' ,, M ..- 4 - .- s B/ml in 0.5-pi drep. I

+ ' .

_ l

=~.-

y w. Q, ;

. s

._ W , ...

w

, .; . 3 .g G - ;- s..

f'ff;,

1

' ,. c 1 s er' .#-. -

- ,M- {

. .. .. . ,. . s s. ,

. p .5 .

..,- y' N

l 3 '. - .

[(d) \ . . _

l self-absorption in the cells. In view of the experimental var. H contents in human tissues. If sample size or exposure tu : '

iables to be encountered in the lysing of cells and the deposi. is inonitored by the hydrogen capture y ray (at 2.23 meV' tion and drying of the resulting slurry on the cellulose nitrate care must be taken to account for varymg contents of H '

fdm. it is clearly desirable to obtain a calibration curve using different tissues. Most importantly, Na content is seen -

boron standards mixed with unboronated cells at cell con. vary from 8% to 18 %, and thus will be respcusible for sign _

centrations equivalent to those used with the unknown slur. icantly different backgrounds. We have takea multiple ren' ry, ings of various " blank" tissue samples fro:n our mouse er

) rat tumor models, and have found that backsiound must t corrected according to tissue type as well as sample size. Il--

example, the background produced by Na from gram que lit. DISCUSSION tities of various tissues varies from the equivalent of ~ 0.5r

. Endogenous boron levels in humans are ~0.1 g B/g in B/g in muscle to ~ 1.1 g B/g for blood or tumor. Th=

l blood, for example; ( Ref. I 8 ) . The lowest level of B consid- the apparent boron content ofeach unknown tissue sampie ?

cred to be useful for therapy is ~ l g B/s tumor. These corrected by subtracting the appropriate B equivalence er levels define the limits o(sensitivity ofinterest for prompt y its Na content as derived from both the tissue type and c analysis of boron in tissue samples, although different re- known sample mass.

quirements may obtain 'n other fields (i.e., plant physiology, Although the relative background obtamed at the MRt l biochemistry, and environmental studies). As described prompt.y facility is ~ 2 times that at the HFBR, the sensitr i above, the two prompt.y facilities at BNL provide this capa. ity and background is such that the analysis of Ipg '?B/--

( bility, with the additional sensitivity at the HFBR used for tissue is possible in a single 200 s run. Therapeutic appix:

i analysis oflow boron concentrations in tissue often obtamed tion of the B(n,a)'Li reaction will require ~ 15 g B/_

with trial compounds synthestzad with natural boron in tumor for an epithermal beam, or ~ 30 g/g for a therm-Since the response from I g of B is only a few percent of beam. It would be hoped that boron 10 levels in norm background, background subtraction is of great importance tissues would be a factor of from 5 to 10 less (i.e.,1.5-6 A in the prompt y method. In particular, the prompt y from B/s). Consequently it has been possible to develop an ost l Na ( 472 kev ) interferes with B analysis and must be correct. line capability for boron analysis that will make possible tr l ed for, to obtain accurate results. Table I summartzes Na and determmation of boron content in blood immediately prw  !

teesmesi Physees, vel 13, Nn.1. Jan/Feb 1tes

&N __ - -

A

-,' ,,c.ae e e, wic :ararym 'cr sorte :y " e s. . . .. ' esct or .

>*C than the determinauen af meanmm! sur. .4. .- ~

mixed fields of neutrons and y rays, which would ef .eune j '

also be accompanied by appropnate controls. The track.

j' _

etching techniques desenbed above provide the capabihty for this type of analysis, as cells may be lysed and deposited

$ on film for boron analysit,.

,', Additionally, needle biopsies may be obtained from tumor j ,,

immediately prior to NCT, for boron analysis by track etch-

,s ing. The whole procedure could be completed within I h,

. thus providing a most important paran .:ter for determining

, , , the requisite exposure time for tumor control. l ooi Qi 1.0 100 i pg B per mt

, irradiated .

C. (a) Bad .

ACKNOWLEDGMENTS f" g cahbranon curve for cells (bottom curve: 0.45 x 10' cella per drop)

1. to'ahbranon curve for standards (no cells, top curve) on cellu-1 drop. y ' We are indebted to Dr. Walter Kane of BNL's Nuclear i

• "'" Energy Department, who served as consultant on the var- l ious neutron physics aspects during the course of these ex- j e, and dunng, irradiation of humans for NCT. Thus one of I pertments. In particular, Dr. Kane designed the Lit CO col- l l tw pnme requisites for future clinical trials has been met.

3

limator used at the MRR, and shown in Fig. 3. We also l g,,thermore, only conventional shielding materials were acknowledge the help given in the early stages of this work

,ud, so that any reactor facility considering NCr should be by Dr. Karmi Ettinger, University of Aberdeen. Scotland, shir to construct a similar facility. and the continuing advice and assistance provided by Dr. )

l'r+sbly the most difBeult way to evaluate a boron com- Lucien Wielopolski and Dr. Daniel Slatkin,' Medical De- l g for potential efBeacy in NCF is to try to measure partment, BNL.

twnor control in ammal tumor models following NCT. This amasaates irradiation of multiple nrumale at multiple dose i paints followed by prolonged observation, with concomitant espenments on controls in the neutron field as well as in x-rn beams. The development ofinformation about the boron

'**d under tract Na DF-AC02 76CH00016 mth the content necessary for successful NCr and abo the poten-signif-

• }"" g ual etreets due to variation in microdistribution *' Preunted in part at the nrst International Symposum on Neutron Cap-Cure time aantly reduces the necessity for such time-consuming and ture Therapy, oct.19:3; Brookhaven Nanonal Laboratory Report No.

83 meV)- 51730.

ripensive e1periments. The basic information is then avail.

" Fuubnght fenow, 1984-85; work supported in part by Deutsche Fors-GD of H in able from the boron content of ammal tissues measured as chunp-gememachafL Se8n 10 Jescribed above. 'W. H. Sweet, in Pr~a-hap of the Ist Internanonal Sympomum on Neu. i for signif. Some basic parameters are more casily obtained from in tron Capture Therapy, BNL Report No. 51730.1983, pp. 3%388.

fple read- arm expenments in cell cultures. The biological activity of 8

K. Kanda, T. Kobayashi, and K. Aob,in Pr~amhap of the 1et laterna- j ucoal Sympomum on Neutron Capture nerapy BNL Report No. 51730. '

Ettse and I boronated antibodies and the binding of melanin affinic O must be ' molecules as well as that of steroid hormone and nucleoside d'.Nian. 0$rownen, and M. Forrut. in Pr~==dmp d the la Inter.

) size. For , analogs may all fit into this category. A simple boron analy- nauonal Sympomum on Neutron Capture Therapy, SNL Report No. ,

am quan. l us on a cell culture of ~ 10' cells (~ l mg) is much simpler 51730,19s),pp.ss-9s. J

'Y. Hayakswa, T. Inada. S. Harana wa. and H. Hatanaka, in Pr~==ha e or P-0.5 g ? the 1st Internanonal Sympomum on Neutron Capture Therapy. BNL Re-W. Thus port No. 51730,1983,pp.77-87.

sample is . 'Prne.= amp of the let Internanonal sympomum on Neutrun Capture f ata l Percent by weght of H and Na in human tusues, from Snyder Therape, edited by R. O. Fairchild and O. L Brownell BNL Report N1

@nce d h j' $1730,1983.

3 and un- 'A. F wr-cyk, J. Menner, and C. E. Pierce. Anal. Chem. 43, 27 I ( 1971 i H Na 'D. N. Slatha, P. L 'dioca. A. Forman. D. Osbel, L Wielopolski, and R.

le MRR T== Wt.(s) s/tisme W:. % s/ tissue Wt.

• O. Fearchild. Biochem. Pharmacol (in ptina).

I sensitiv. - 'M. P. Failey, D. L Anderson, W. H. Zouer, O. E. Gordon. and R. M  !

, arma 1400 15/ 3.7 5 0.1s man, cat, Anal Chem. St. 2209 (1979). i

~I sog

• Blood 5 500 550 10.0 10 0.1 *E. S. Gladary, B. T. Jurney, and D. B. Curus Anal Chem. 4a. 21 M ]

. applica- 5eallistesone 640 67 10.5 0.64 0.10 (1976).

i rg 80B/g Imr 1800 180 10 1.8 0.10 '*T. Kobeyenhi and K. Kaada, Nuct Instrum. Methods 204, 523 ( 19s b

) thermaj Muscle 28 000 2300 10 21 0.08 "R. L Rascher and P. B. Pnce, Science 140, 1221 (1 % 3).

Pancries 100 9.7 9.7 0.14 0.14 "M. Ldental, Anal Chem. M 1270 (1972).

)nonng 1000 99 9.9 1.8 0.ls "M. Thenaar. C. Heurtanus, and L C. W'meocq, Brata Research 19e l'5 Luns 3.5-6 g Pariachyma $70 55 1.0 09:03, Ihad 430 44 0.3 "B. Larsson, D. OsbeL and H. O. Barner, Phys. MacL Biol 29. 34 I ( i osa i l 3 an on.

n'hje the $#ue 180 18 10 0.22 0.12 "D. Unbid.1. Hocke, and W. Elsen, Phys Med. Boot 28.1453 ( 19s h

"" 11 2 '*R. O. Faarchild and V. P. Bond, lat. I Radiat. Oncot Boot Phys i1. t )I b rior

~y p (1985).

1 Physics.Vol 13, No,1.Jen/PeB 1988

04 FaercNd at st. Wicroscarysis 'or .icron my +e %as .. eac e '

f.NI

) ' t; 'D Gaeel. R G Fur:nud. H G Bormer. and B Laruon. Radiat Res 98. **

307 (1984)  ;> mum en Neutron Captare Therapy. BNL Re;w:rt %) .q b W. S. Snyder, Chairman, Report of the Task Gmup on Reference Man, 120-127.

4 i f' / CAP Aeperr No. JJ (Pergamon, New York.1975).  ;"D. Gabel, R. G. Furchdd, B. Larsson, K. Drescher, and W R Rg Pr*mp of the 1st International Sympossum Neutron Capcurt l '7. Kobayashi and K. Kanda, in Prar==diny of the ist International Sym.

apy. BNL Report No. 51730,1983. pp.128-133.

l l

l

)

)

I

(

i l

i k

i h

I l i

l 1

l l

i Meesel Physion, Vol 13, No.1. Jan/Feb 1985 g ..

1 l

)

i (  !

DEPARTMENT OF HEALTH & HUM AN SERVICES Public Health Service

.%',,,,,, NationalInstitutes of Health Memorandum 0:ts l Frcm Subject f

To

! Experimental protocols for irradiation experiments of Oct. 21st and 28th in AFFRI.

Assuming a positive response on the part of the safety committee meeting on the a.m. of the 19th OCT., with regard to the use of cellulose nitrate film, the plan for Friday 21st of Oct. is as follows:

We will undertake 7 runs.

7 The following are the desired fluences:

1 x 10 12 n/cm 2 5 x 10 12 n/cm 2

t. I x 10 13 n/cm 2 These will be undertaken twice, once in the Carmichael set-up,and once in the Loughlin/ Matson set up of the 8th and 15th of Dec. 1987, using either 9" D2 0 and 0" H2 O or 3" D2 0 and 3" H2 O (to be discussed with Captn. Manson)

The seventh run is intended to provide a worst case sample' for i safety and should therefore be run at the highest fluence, and should probably be run first.

) tissue slices 10 microns thick, contain-ThefilmwillbecoatedwigBenrichedBo,knowntoproduceappropriate ing a uniform quantity of pits or etchings on the film.

The film will be mounted in either cardboard or plastic card mailers, which should hold it upright in the beam, and which will be tailored to suit the appropriate set up.

The experiment is designed to examine the system and to check that ,

the fluence is uniformly distributed in both designs, to identify l the optimum fluence for irradiation, and to subsequently develop the film and devise an optimal numerical system for quantification of pit density.

, i a

E I

I The experiment planned for Friday Oct 28th will utilize the optimum design and fluence of those used the previous week.

We will uniformly irradiate tissues impregnated with varying concentrations of Bo compound, in order to devise a dose-response curve I W kmck "N - __ __ _.

to]'

We would anticipate six-eight runs.

The optimum fluence-and design will be selected on the basis of the previous weeks experiments.

I I

I I

I I

I I

I I

I I

I November 2nd will be devoted to cell irradiation, assuming the I incubator has been installed in the prep. area.

Two tumors will have been excised by surgery, cultured, and will be irradiated in the presence of the Boron compound bound to the appropriate hypothalamic factor.

We will need four runs at 500 watts x 20 mins, and four other runs, 2 at 500 watts for 5 mins and 2 at 2KW for five mins.

I -

E I

I I

I I

g i LI I ,

I I