ML20214M926

Suppl B of FSAR on Nbs Reactor

ML20214M926
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Site:	National Bureau of Standards Reactor
Issue date:	12/16/1966
From:	NATIONAL INSTITUTE OF STANDARDS & TECHNOLOGY (FORMERL
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ML20214M923	List: ML20214M920 ML20214M926 ML20214M929 ML20214M932 ML20214M938 ML20214M946
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166, 9464, NUDOCS 8609150035
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NATIONAL BUREAU OF STANDARDS REPORT LI

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FINAL SAFETY ANALYSIS REPORT ON THE l

NATIONAL BUREAU OF STANDARDS REACTOR ll NBSR 9B l

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I U.S. DEPARTMENT OF COMMERCE '%MP NATIONAL BUREAU OF STANDARDS EGU_ATORY J0CXET f!LE COPY 8609150035 661221 PDR ADOCK 05000184 or

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THE NATIONAL BUREAU OF STANDARDS I

The National Bureau of Standards is a principal focal point in the Federal Government for assur.

ing maximum application of the physical and engineering sciences to the advancement of technology in industry and commerce. Its responsibilities include development and maintenance of the national standards of measurement, and the provisions of means for making measurements consistent with those standards; determination of physical constants and properties of materials; development of methods for testing materials, mechanisms, and structures, and making such tests as may be neces-sary, particularly for government agencies; cooperation in the establishment of standard practices for incorporation in codes and specifications; advisory service to government agencies on scientific l and technical problems; invention and development'of devices to serve speci~al needs of the Govern- E ment; assistance to industry, business, and consumers in the development and acceptance of com.

mercial standards and simplified trade practice recommendations; administration of programs in cooperation with United States business groups and standards organizations for the development of international standards of practice; and maintenance of a clearinghouse for the collection and dissemination of scientific, technical, and engineering information. The scope of the Bureau's activities is suggested in the following listing of its three Institutes and their organisational units.

Institute for Basic Standards. Applied Mathematics. Electricity. Metrology. Mechanics. Heat.

Atomic Physics. Physical Chemistry. Laboratory Astrophysics.* Radiation Physics. Radio Standards Laboratory:' Radio Standards Physics; Radio Standards Engineering. Office of Standard Reference Data.

E Institute for Materials Research. Analytical Chemistry. Polymers. Metallurgy. Inorganic Mate- 3 l rials. Reactor Radiations. Cryogenica.* Materials Evaluation Laboratory. OfEce of Standard Refer.

ence Materials.

Institute for Applied Technology. Building Research. Information Technology. Performance Test Development. Electronic Instrumentation. Textile and Apparel Technolog.y Center. Technical Analysis. OfEce of Weights and Measures. OfEce of Engineering Standards. OfEce of Invention and Innovation. Office of Technical Resources. Clearinghouse for Federal Scientific and Technical e Information.*

• I *tocated at Boulder, Colorsde 8 I

" located at 5285 Port Royel Ao.0301.

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NATIONAL BUREAU OF STANDARDS REPORT I NBS PROJECT NBS REPORT 3140100 9464 l

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I of the Final Safety Analysis Report on the National Bureau of Standards Reactor NBSR 9B December 16, 1966 I

I IMPORTANT NOTICE N Afl0 N AL BURE AU OF ST AND A R DS REPORTS are usually p eliminary or progress accounting documents in' ended for use within the Covernment. Before matenalin the reports is for.mally pubbshed it is sub ected i to addrtional evafast+on I

and review. For tnis reason, the publication reprinting, reproduct on. or open literature listing cf this Pe; ort, either in whole or in part, as not authorized unless permissio$ :s obtained in writing from the Office of the Director, Nat onat

' Bureau of Standards. Washing'on. D C. 20234. Such permission is not needed, homever, by the Government agency for .5;ch the Report has bee 1 specifically prepared i' t: sat agency wishes to reprodace additicial copies for its own use.

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U.S. DEPARTMENT OF COMMERCE I NATIONAL BUREAU 0! STANDARDS t >

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PREFACE The letter of December 7,1966 in the Docket No. 50-184 for the NBSR has requested additional information in a series of nineteen questions. The following Supplement B to the report No. 8998 entitled NBSR 9, Final Safety Analysis Report on the National Bureau of Standards Reactor contains the re-sponse to these questions. Each response is preceeded by the original question of the above mentioned letter.

In addition, revisions to various sections of NBSR-9 are included.

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I I TABLE OF CONTENTS

1. Expand the discussion in your October 1, 1966, submittal to include 1-1 I the results of tests conducted to determine the efficiency of filters in the emergency exhaust and emergency recirculation systems. Pro-vide a description of the iodine filters including sufficient physical, chemical, and m'echanical properties to permit an independen't estimate I to be made of the potential heating of the filters in event of an accident.

2. Describe the program established to demonstrate the reliability of 2-1 I the dynamic protection system for the reactor confinement building, including the schedule being followed and any results obtained to date.

3. It is our understanding that the thickness of the elastomer coating 3-1 painted on the internal valls of the confinement building does not I conform to your description in the FHSR. Provide a discussion of the situation.

4. With respect to the underground emergency control station: 4-1 (a) Discuss how the potential for negating the function of the confinement building by maloperation of the controls I has been eliminated.

(t) Discuss provisions made to make the station habitable.

What assurance will an operator have that occupancy I within the station following an accident will not jeopardize his safety?

5. It is our understanding from informal discussions that significant 5-1 modifications are being made so that the confinement building control system will meet the following criteria:

(a) All controls will be redundant up to the motor starters or final solenoid valve for each individual component.

(b) All sensing instruments (radiation and pressure detectors) and master relays will be duplicate and independent.

(c) Sufficient indication and control will be availabic in the emergency control center to indicate the status of each vital component and to allow manual control should its I autenttic control malfunction. This nanual backup shall be in additi:n to the redundant autca stic controls.

ProvLda a description of ti.e modifications being n2de supported by drawin;s of the codified system.

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I Question 6.

Page (a) Expand the discussion in your October 1, 1966, submittal regarding 6-1 I the capability of the confinement building to perform its design function with some degree of failure in the sealing of the building to include the failure of automatic closure valves and building I personnel doors. Sumarize the results to allow identification of the items which must function, the backup device for each item, and the consequences'should it fail to operate.

I (b) As a result of informal discussions we understand you are planning to make the following modifications:

I (1) Provide indication of the open-closed status of the confinement building doors in the emergency control station,as well as in the control room, and to pro-I vide alarms at both locations whenever a door is open for more than a selected time.

(2) Take suction for the emergency exhaust system from I areas of lowest expected contamination in event of an accident.

I (3) Have the stack dilution fan shut off during building isolation.

7. (a) Discuss the additional personnel and engineering support services 7-1 1 to be available for initial startup and operation. Provide an organizational chart showing the relationships of these personnel and services to each other and to management. Discuss the adequacy of this startup and initial power organization.

(b) Describe the composition of the Reactor Safety Comittee, its I charter and organizational position, and the criteria established to insure that a portion of the membership will be independent of any responsibility for operation, use, or management of the NBSR.

8. Your October 1, 1966, reply to our previous question regarding 8-1 the maximum reactivity insertion whose energy could be contained without rupturing the reactor vessel or primary system should be expanded to include transient as well as steady-state considerations. If, as it appears, the vessel will be contained from gross strain failure by the backup biological shield, the restraining effect of this shield should I be considered. Similiarly, failure of the beam tube nozzios should be considered, and the ef fect of their failure should be factorei into the analysis. Also, consider the ef fe:ts of vessel or top plug ;uap.

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9. Your October 1, 1966, reply to our previous question regara.ag the failure of a shim arm should be expanded to include consideration of 91 I the consequences of the maximum reactivity transient which could occur under any rod programming scheme. These consequences should be compared with the energy absorbing capability of the reactor vessel. If the vessel cannot contain the transient, you should also include consider-I ations of the confinement building'c capability to contain or limit the consequences of the excursion. If, as you indicated in your previous reply, it is necescary to provide means to assure reactivities do not I exceed acceptabic values you should include your analysis of more posi-tive means than administrative control. For exampic, consideration could be given to combinations of such schemes as:

(a) rod programming interlocks; (b) rod steps which prevent a broken rod's falling; (c) negative period scrams; I (d) reduced control rod worths; (c) any others.

Provide complete details on any schemes employed.

10. Provide an analysis of the tritium doses resulting from a massive heat 10-1 exchanger failure in which the heavy water carries over to the secondary I cooling tower.

11. In your answer to our previous questions, you indicated that in situ 11-1 I tests of the reactor safety system at the maximum control room tem-peratures you would expect to operate were not planned. Although factory tests were conducted it is our opinion that elevated temper-ature tests should be repeated to detect possibic cable faults, cold solder joints, differences in component cooling arrangements, proxirf ty of heat producing equipment, etc. Please infom us in detail of your reasons for not performing such tests if this is still your intent.

12. Provide the following infomation regarding the electrical distribution 12-1 system:

(a) llow long can the emergency battery supply emergency loads at its minimum acceptabic charge condition? What loads are involved?

(b) llow is switchacer operating power supplied and is this power I split such that single failures will not violate the split bus conceptt I

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I Question 13.

Page It is our understanding that all potentially radioactive liquid waste 13-1 I will be batch sampled prior to discharge. Provide sufficient details, including any changes to the system which may be required to allow us to evalut.te this approach.

14. Provide the following information regarding the Emergency Cooling 14-1 System:

(a) The maximum and minimum flow rates from the system when the bottom feed method is employed.

I (b) A quantitiative analysis of the ability of the top feed method to adequately cool the fuci elements.

I -(c) Modifications required to the system to assist the operator in his determination of whether to use the top or bottom feed method.

15. Provide information regarding testing that was conducted to verify the 15-1 operability of the reactor vessel fuel handling equipment.

I 16. In order to' establish limits on the number of fuel elements which can be handled at any one time, provide the assumptions and results of a calculation shouing the minimum number of fuel elements which could 16-1 achieve criticality in any geometry, under any conditions of noder-I ation. Discuss the provisions taken to assure that such geometries could not occur in the event of fire in the fuel storage vcult,

17. Provide a list of manually operated valves whose incorrect positioning 17-1 could have 4. significant effect on safety. What action will be taken to ensure that these valves are positioned correctly?

18. Expand your October 1,19f 6, reply to our previous question regarding 18-1 mechanisms for pressure differential in the confinement building to include:

(a) The extent of Icakage which might be expected in the first two hours. Include considerations of heat sinks if credit for those sinks can be quantitatively proven.

(b) The extent of Icakage over the course of the accident, listing your assumptions and justifications thereof.

(c) Calcula:cd two haar and loa,, tern dosca. List all aasumptions and ba m therefer.

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I Question _Page Your calculations should include considerations of decay heat for the course of the accident, both with and without thermal shield cooling.

For the long term calculations the total heat content of the primary l

coolant and plant components must be accommodated. Solar " pumping",

including possibic vacuum conditions, should be considered.

I lour calculations should be sufficiently detailed to allow a deter-mination of the adequacy of the emergency exhaust system sizing, the filter design, and the building integrity.

19. We note that the IGSR is in an area of low scismic probability; however,19-1 it is still necessary to analyze quantitatively the potential conse-quences of an earthquake of maximum potential magnitude for the ' area.

I Include both the confinement building and the reactor plant in your considerations.

Cencral Revisions Void Shutdown System 20-1 I Spent Fuct Storage Facilities 21 1 Shim Ann Shaf t 22 1 I

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I QUESTION 1.

4 Expand the discussion in your October 1, 1966, submittal to in-1 clude the results of tests conducted to determine the efticiency of filters in the emergency exhaust and emergency recirculation systems. Provide a description of the iodine filters including sufficient physical, chemical, and mechanical properties to per-mit an independent estimate to be made of the potential heating of the filters in event of an accident.

i RESPONSE 1.1 ABSOLUTE FILTERS The table below gives the results of the in-place tests of the absolute filters in exhaust and recirculating fan systems.

The tests were performed by release of dioctyl phitolate (DOP) acrosol upstream of the filters and sampling upstream and downstream of the filters.

TEST DATE FAN EFFICIENCY 11/1/66 Reactor Basement Exhaust EF-27 99.95?.

11/1/66 Recirculating System SF-19 99.937, 11/1/66 Normal Exhaust System EF-3 99.967.

12/7/66 Emergency Exhaust System EF-5 99.9667.

I 12///66 8/6/66 Emergency Exhaust System EF-6 Irradiated Air Exhaust Systein EF-4 99.9717.

99.967.

I "Since the air-operated generators produce a polydisperse aerosol with the average particle slightly larger than 0.3 p, the in-place test is not intended to eliminate the desirability of has ing new high-ef ficiency filters tested according to Instruction I Manual 136-300-175A (Oak Ridge Test Facility Testing).

ference of 0.027. is allowed between the rating of new filters by A dif-the Quality Assurance Test and the rating of filter systems (in-I cluding single installed f11ters) by the in-place test. To qualify as high-efticiency, the system or installed filter must have an efficiency of 99.957 in the in-place test."I The filters for SF-19 were lower than the 99.957. Icvel; how-ever, this fan is not in an exhaust system, being used only for internal recirculation during building closure conditions. Filters I for the exhaust of lab, C-Ont and c.002 has ho au m of ai t ti cal Foa ! i n ., r i l l a t i o n .

no* Leen installed t' p u n ccrpletion the w 1

G. W. Fellholtz, " Filters Sorbents and Air Cleaning Systems as Engineered Safeguards in Nuclear Installations" ORNL-NSIC-13, October l%6, page Ill.

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h filters will be tested and accepted in accordance with the above criteria.

1 1.2 IODINE FILTERS L t The activated charcoal filters in the Recirculating System r SF-19 and in the Emergency Exhaust Systems EF-5 and EF-6 were tested for Iodine removal on November 22 and 23, 1966.

• The tests were performed by release of I plus elem(ntal l iodine carrier before the filter with sampling fore and aft of the bank, and iodine determination by comparison of fore and af t sampler counting rates. This method is direct, requires na radio-P chemical manipulations,-and constitutes a hazard ~ 1000 times less L than that obtaining to 1131 The principal disadvantages are re-toted to its short half life (25 min.), the need to employ the sources shortly after their generation, and to correlate measure-p ments with a stop watch.

L, One gram elemental iodine sampics were weighed into small I quartz vials which were fused shut with a breakable tip. These were bombarded in a flux ~ S x 1012 n/cm2 sec as needed and for times varying from 10 sec.

to 100 sec. to produce sources varying F~

f rom 1 mci to ~ 5 mci when used.

L The sources were bombarded at the NRL reactor, transported in lead pigs to NBS and installed in the " duct vapor generator" at the start of a test. The " generator" consisted of a container which roccived the quartz vial and allowed the quartz tip to be broken after containment. The generator had connected to it stainless steel tubing which entered the duct and allowed the iodine to emmanate from numerous small holes. Sublimation and iodine transport was effected by passage of warm (~ 700 C) air F through the stainless steel tubing-generator. Iodine release L took place several feet up stream of the filters.

Ai sampling was of fected by devices which drew approximately 1 CFM through 957. of the iodine 2"sampled.

diameter f11ter pads that absorbed approximately These pads (fore and af t) constituted the samples that were subsequently detected by s-counting and (when sufficiently active) by y-counting.

k EF-5,Two EF-6,separate tests were accomplished on each of the filters, and SF-19.

e The S-counter was approximately 5 times more sensitive than the y-counters thus, in some of the cases only S-counting results were obtained. The results are given in the table below. In all cases of clearly detectable 1128 '

the decay followed a 25 min, half life, l

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I FAN EFFICIENCY ,

I EF-5 99.9967.

EF-6 99.9947, 3F-19 99.9977.

The physical dimensions and charcoal content are shown in I Figure 1.1. A thermal conductivity for the charcoal under non-flowing conditions can be estimated to 207. accuracy using the technique described by Etherington 2 where the thermal conductivity ratio of the charcoal to air for packed beds filled with a stagnant

,I fluid is considered. Knowing the fractional void volume which is approximately 757. for NBbR filters and the k to k ratio a thermal conductivity for the mixture of 0. bib Nib /hr f!h F/f t was calculated.

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I I QUESTION 2.

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Describe the program established to demonstrate the reliability of the dynamic protection system for the reactor confinement building, including the schedule being followed and any results obtained to date.

I 2.1 TEST PROGRAtt A periodic test program has been initiated to test the re-liability of the building closure system routinely twice a week.

I This test f requency will prevait until initial fuel loading, then revert to approximately once per month.

I The test is designed to test every function of the building closure system. Initiation by the various methods, (1) radiation nonitors RD-3-4, RD-3-5, or RD-4-1 and (2) the manual switch, S-3 (see Figure 5.1 of this amendment), is alternated on each test.

I The system is tested in the "as is" condition.

After initiation the pressure control systems automatically reduce building pressure to -0.25" 110 2 by operation of the ener-gency exhaust fans. Proper indication and control functions of all equipment is checked from the emergency fan panel and the I control room'. After normal operation is observed various faults are simulated to check the backup features of the standby exhaust fan. This is accomplished by turning the AC control switch "off" I for EF-5 and introducing a leah. The standby fan EF-6 AC is then checked to see that it operates to maintain the building at a negative pressure. After returning the system to normal emer-gency operation (EF-5 AC in " auto", EF-6 AC in " standby" and the induced Icak closed, a loss of AC power is simulated by turning of f AC breakers to EF-5 and 6. Proper operation of the DC " auto" and " standby" feature s are then checked for both EF-5 and 6 as I was performed on the AC controls described above.

features of the major scram FSR and DSR relays are also checked by visual inspection. llaving completed the test, the closure Redundant system is returned to the ready position.

_2 . 2 RESUI,TS I Tests have been conducted a number of times with successful resulta on all parts of the systen that were operational at the time of each test.

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of a manual major scram to isolate the reactor building and the pressure controllers automatically reduced building pressures.

The number of closures for these purposes approximate six with a minimum of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> sustained operation of the automatic pres-sure control functions. Due to modifications in the ventilation p system for labs C-001 and C-002, involving EF-23, in progress at L this time, that portion of the test has not been checked to date; however, in all tests performed to date the doors at all entrances to the confinement building closed and scaled properly, all valves functioned as designed, the emergency exhaust fan functioned and the internal recirculating fan performed properly. System in-dications in the control room functioned. There were some errors in proper light indications at the emergency fan panel, caused

{ from construction installation in labs C-001 and C-002, but these have now been corrected.

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QUESTION 3.

It is our understanding that the thickness of the elastomer coating painted on the internal walls of the confinement building does not confom to your description in the FilSR. Provide a discussion of the situation.

RESPONSE

Though the containment building well exceeds its tightness specification (by a factor of five), certain areas within the containment structure were not given the full thickness of

{ clastomeric coatings called for in the painting contract. This was shown to be the case by two independent investigations. One was performed by the Moore Research Laboratories, engaged by the General Services Administration; and the other by the Materials

- and Composition Section of the NBS Building Research Division.

- At the time of this writing, both the General Contractor and the painting Sub-Contractor have admitted to these contractural deficiencies and have further.nore agreed to make the necessary a repairs. It only remains to ef fect a detailed arrangement ,

'between NBS, GSA, and the contractors. The work should begin in the very near future and when completed will fully comply with the clastome ric thickness stated in 3.3 of NBSR 9,

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- Reactor. In the perforinance of this work the Moore Research Laboratories will act as Overriding Supervisor for both Nbs and the GSA.

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u With respect to the underground emergency control station:

[ (a) Discuss how the potential for negating the function of the confinement building by maloperation of the controls has been climinated.

(b) Discuss prdvisions made to make the station habitable.

What assurance will an operator have that occupancy within the station following an accident will not jeopardize his safety?

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RESPONSE

r' L 4(a) The operator has direct control of the following valves and fans from the emergency fan control panels p 1) Vacuum breaker valve ACV-12,

% 2) Recirculation fan SF-19,

3) Emergency exhaust fans EF-5 and 6,

4) Process room exhaust valve ACV-10, and

5) Dilution fan EF-2.

Only two of these five control functions a f fect confinement integrity. These arc (1) the vacuum breaker valve, ACV-12 and (2) the emergency exhaust fans EF-5 and 6. ACV-12 is a 4 inch butterfly vacuum breaker, installed to protect the reactor building from externally applied pressures or internal building vacuum, the mechanisms and magnitude of which are discussed in Response No. 18

( of this amendment. Normally valve operation is automatic as de-L scribed in NBSP, 9A, Responses No. 5 and 12. This vacuum breaker was installed primarily to protect the building from damage as a result of fan and/or valve malfunction during normal ventilation

{ conditions. Although considered incredibic because of redundant functions to stop normal ventilation fans in major scram conditions, manual backup control of this valve has been afforded the operator, who aided by valve position lights and internal building pressure indication is able to monitor ACV-12 operation if opening should be required.

Operation of EF-5 and 6, emergency exhaust fans, causes their respective closure valves to open and close as cach f an is operated. Normally this in an automatic operation s however, the operator can assume manual control of fan operation only. Ito is not able to opon or close EF-5 and 6 dischar.:e and/or suction i

valver, en tpt A:'/ 10 ( we r!curo i.16 of' thi. n nd ent), in.

dipenAnt et tan uprratlen.

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I 4(b) The emergency fan control pit as built was not provided with space heating, subsequently electric heaters and a dehumidifier have been installed to maintain habitable conditions.

The present location of the emergency control station was chosen to provide an environment in the event of a major core meltdown which would be well shicided from radiation escaping the confinement buildind. Examination of the building structure, as schematically indicated in the Figure 4.1 below, shows that rhe minimum thickness of earth interposed between a person at I the emergency station and the confinement building is more than 40 feet. In view of this the major source of radiation to a person in the vault would probably be due to scattered radiation I entering the top hatch of the vault. A very simplified con-servative calculation was made to estimate this effect.

assumptions concerning the gamcna ray swrce were the same as The previously used, namely 1007. release of Xenon and Krypton and I 507. release of iodine following 10 MW operation for 180 days.

Half of the building source was considered to be at the center of the upper level of the confinement building where the wall I thickness is a minimum of 1 foot. The average energy of gamma rays emitted by the source were taken as 1 Mev. Although credit is taken for the concrete absorption no gamma spectrum shift was evaluated. All gamma rays which reach the hatch are I assummed to scatter by Compton scattering. All intervening structure is neglected. The dose rate is calculated to be 0.6 mr/hr averaged over the first hour.

It should also be pointed out that the control station is otherwise habitabic and readily accessible. Routine inspection I of the vault is to be expected during each operating shift by reactor operating personnel.

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QUESTION 5 l

It is our understanding from informal discussions that significant I modifications are oeing made so that the confinement building control system will meet the following criteria:

I (a) All controls will be redundant up to the motor starters or final solenoid value for each individual component.

(b) All sensing instruments (radiation and pressure detectors) and master relays will be duplicate and independent.

l (c) Sufficient indication and control will be available in the emergency control center to indicate the status of each I vital component and to allow manual control should its automatic control malfunction. This manual backup shall

• be in addition to the redundant automatic controls.

Provide a description of the modifications being made supported by l drawings of the modified system.

RESPONSE

1 The confinement closure system has been revised in accordance I with the attached drawings. Figure 5.1 shows the added major scram relay and the additional radiation monitor RD 4-1. This arrangement is designed to provide redundant radiation monitor scrams on high I activity in the exhaust syster..s, RD 3-4 in the irradiated air system, RD 3-5 in the nonnal exhaust air system, and RD 4-1 in the stack exhaust system. Should RD 3-4 or RD 3-5 fail, RD 4-1 nonitors l

I their exhaust and would be available for detection. Contacts for both K-114A and K-114B have been placed on each side of coils for the FSR and DSR relays, thus eliminating the probability of the relays remaining energized with a single fault. These circuits a re fed power f rom DC panel DCP-2, circuit #3 (See NBSR 9, Figure 9.7 for the integrated DC electrical system which feeds DCP-2.). Ifanual initi * '.on of a major scram can be ef fected with switch S-3.

Figures 5.2 thru 5.4 show the added FSR contacts which isolate the control relay for SF-3, SF-1, and SF-12 f ans. There are no closure valves associated with these fans.

I d raw in,;s , fan SF-3 has contacts f rom FSR-l and FSR-2, either of which when de-energized will cause the fan to stop.

As indicated on the Fan SF-1 has contacts from FSR-2 and FSR-3 and SF-12 has contacts from FSR-1 and FSR-2. (The switchy; ear cubicle and motor control center is given on each drawing. Their interconnection to the electrical .

distribution system is given in Figure 3.12 of N13SR 9.)

Tim elec t ric if tnd pn. ur it ic cont rols for faq SF-2 L ive been r nc i ' s ! F. F i ci r. .3 ind ".h. O! !it ic nal E SR c .m t a c t . Eno b t n 10 t t i 1. ! . .o: i n ' . i' iiet . tit ' . F T '. - 3 ur.  :- 2. Pa . r 5-! .

I feed to the solenoid valve for ACV 1 passes through the added FSR-2 contacts, thus assuring that the solenoid is de-energized i by relays M-44 or FSR-2.

Similar revisions have been effected to SF-11 and EF-27 I electrical and pneumatic circuits as shown in Figures 5.7, 5.8, and 5.9. Additional FSR contacts have been added.

Two sets of contacts from FSR-3 and DSR-1 have been provided to give duplicate auto operation of SF-19, recirculation fan, shcwn in Figure 5.10. A manual "on-off" switch located on the emergency I fan panel permits remote manual operation should the automatic features fail. Oversize electrical overloads have been installed to prevent undesired trips and yet protect the electrical distribution system from a possible short.

Fan EF-3 shown in Figures 5.11 and 5.12 has been revised by I feeding solenoid valve power through an added DSR-1 ccntact. Revisions to the control board switch assure the fan will shut off on a major scram with the switch in either " hand" or " auto".

I Revisions to fan EF-23 are similar to those made to SF-3. Figure 5.13 reflects the changes which incorporate an additional DSR-3 contact in the starting circuit. Figures 3.14 and 5.15 show the added FSR-1 I contact and the relocation of ACV-6 power feed in the electric and pneumatic controls for EF-4.

The control circuits for EF-5 and 6 have been revised in accordance with Figures 5.16 thru 5.21. These revisions (1) remove ACV-10 from automatic opening as discussed in Response # 6 to this Amendment, (2) change the sensing point for PS-150 to make this channel have I redundant control functions with PC-150 and (3) provide redundant auto operation of either fan backed up by manual control at the emergency fan control panel.

Revisions to the building closure doors are discussed in Response

1. 6 of this Amendment.

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I I QUESTION 6.

(a) Expand the discussion in your October 1, 1966, submittal re-garding the capability of the confinement building tc perform its design function with some degree of failure in the scaling of the building to include the failure of automatic closure valves and building personnel doors. Summarize the results ,

to allow identification of the items which must function, the backup device for each item, and the consequences should it fail to operate.

(b) As a result of informal discussions we understand you are planning to make the following modifications:

1) Provide indication of the open-closed status of the l

ccnfinement building doors in the emergency control station, as well as in the control room, and to pro-I vide alarms at both locations whenever a door is open for more than a selected time.

2) Take suction for the emergency exhaust system from areas of lowest expected contamination in event of an accident.

3) Have the stack dilution fan shut off during building I isolation.

RESPONSE

,I 6(a) In December, 1966, tests were conducted to determine the consequences of failures in the confinement closure system. These

~

tests were similar to those performed earlier and reported in I NBSR 9A, Response No. 5. During the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test with a scaled building, described in Response No. 5 of NB5R 9A, only one of the emergency exhaust fans was required to operate approximately 1/25 I of the time to maintain the building at the -0.25" H 2O control point, thus allowing considerable margin for other effects.

During the December tests the effects of each of the door seals I and ACV valve failures were observed. Building pressure and exhaust flow rates were recorded under various modes of failures and are listed below. Pressure measurements were made with an I

inclined mancmeter sensing differential pressures between the reactor mezzanine and cold lab basement levels. The internal recirculation fan, SF-19 was operated continuously to equalize internal pressures in the reactor building.

The test was performed by initiating a manual major scram signal and allowing the emergency exhaust fans to autcmatically reduce building pre ssun to .23" H3 0 Doer seals were t5en d-flated and 'alv2- c

• n d . c r.e at a t ine . The pres'ure tot  !

n s t e- tWn at te - u &1J the luildin; preszure at -0.'

5 I .

m.:

i.

i .

I H 2 O by operating EF-5 and 6. After the pressure had stabilized 4 I and average flow rates measured, the seals and valves were re-closed. The building pressure was allowed to return to the

-0.25" H2O between steps of the test. Those steps identified by I (1) were performed with EF-5 controlled at -0.25" H 0.

vacuum dropped to approximately -0.10" H 2 O and below, EF-6 would operate. All other steps were performed with both EF-5 and 6 2 When the controlled at the -0.25" H 2O setpoint.

Based on the measured exhaust flows, an exhaust rate ex-pressed as per cent of building volume (600,000 ft )3for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> was determined and is listed in the Tabic below.

EXHAUST STABILIZED EXHAUST RATE I BLDC. FLOW  % OF PRESSURE RATE BLDG, PENETRATION ("H2O) (CFM) VOL/2HR ACV-1 -0.04 300 6.0*

Truck Door (I) -0.03 260 5.2*

ACV-2 -0.06 260 5.2*

SE & NE Personnel Doors Vestibule Drs. Open -0.10 250 5.0*

Vestibule Drs. Closed -0.15. 240 4.8*

SE Personnel Door-I Vestibule Drs. Open Vestibule Drs. Closad

-0.13

-0.21 250 218 5.0*

4.4*

NE Personnel Door I Vestibule Drs. Open(I) -0.10 223 4.5*

Vestibule Drs. Closed (l) -0.17 130 2.6 I

• These exhaust rates exceed the capacity of the two emer-gency exhaust fans which is approximately 6%/2 hours when the additional 2.5'/./2 hours (see Response No. 18 of this amendment) is I applied; therefore, a retest was performed to measure the building pressure with only one fan operating. Tests were conducted with ACV-1 and the truck door seal, the most severe conditions. The I building pressure stabilized at -0.015" H 2O with ACV-1 open and one fan operating and -0.02" U 2 0 with the truck door seal deflated and one fan operating. It was therefore concluded that one fan is capabic of maintaining the building at a negative pressure with I a singic failure while the second fan can exhaust the potential pressure rise from other sources (2.5'/./2 hours). In no case can the actual exhaust rate exceed 6%/2 hours; therefore, the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> I dose for these conditions would be based on this naximum exhaunt rate. (Re fer to Peapon e 'o. 18 of this ,nendment.)

1 I 6-2 e .

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EXHAUST STABILIZED EXHAUST RATE I BLDG. FLOW  % OF PRESSURE RATE BLDG.

PENETRATION ("H20) (CFM) VOL/2HR Elevator Personnel Door Vestibule Drs. Open(l) -0.10 144 2.9 ,

Vestibule Drs. Closed (l) -0.10 125 2.5 ,

ACV-3(1) -0.15 130 2.6 ACV-7(1) -0.15 130 2.6 ACV-6(1) -0.25 120 2.4 Emergency Personnel Exit -

Stairway Drs. Open(l) -0.22 112 2.2 ACV-12(1) -0.23 112 2.2 It should be noted that building pressure and exhaust flow ,

rates are sensitive to the location of an induced leak. This is to be expected and shown in the measured pressures and flow rates I for ACV-1, ACV-2 and the truck door. The exhaust rates are the same er larger, for ACV-1 and 2, however the building vacuum is -

larger than that for the truck door seal failure. Leaks at ACV-1 and 2 bring air into the normal building ventilation duct system

(

thus inhibiting the equalization with the remainder of the building prior to exhaust by EF-5 and 6, while a leak at the truck door is I not restrained by duct work allowing inleakage to reduce the in-terior vacuum more noticeably.

Should a door seal fail on all but the truck door, in-leakage is reduced by closure of the vestibule doors. Rupture of the emer-gency exit seal results in the smallest amount of in-leakage. No backup is afforded the ACV valves, however, each is equipped with a positive closure mechanism as described in NBSR 9A, Response #12.

In conclusion it has been shown that the failure of ACV-1, the truck door seal and ACV-2 are the most severe; however, the exhaust I fans were able to maintain the building at a negative pressure thus preventing uncontrolled out-Icakage.

i I 6(b) 1) It would appear that a more thorough discussion of the NBSR philosophy relative to confinement building personnel door operation is required. From the initial planning of the facility the NBSR staff has felt that ready access to the I research areas of the confinement building was necessary. The number of experimenters that occupy the laboratories and I utilize the reactor will be greater than one hundred. The traf fic betwen the confinement building and the laboratories which ,! ! ! invels both equi; rent and personnel viil l be very ext.n-isi. T.i t , statt nt la r.i!- ca tL hasir of ' M staff I 6-3 e .

O ,

I experience at Brookhaven Nation *al Laboratory as well as Argonne, Saclay, Westinghouse and other laboratories. The

NBSR confinement building is a large building with several

'large penetrations through which ventilation air must pass E

during normal operation. The building is capable of being closed to a degree that makes exchange of air with the out-side atmosphere a very slou process (of the order of more I than one hour to equalize a pressure differential). This is done to minimize the necessity of leaking fission products-in the event of an accident. Under normal operation of the emergency closure features it has been shown that downwind dosage due to fission product inhalation by off-site personnel is extremely low. It has also been shown that numerous of the individual items of this system can fail to operate and provided the remainder of the system functions the dosage I levels are still not high.

t I The NBSR staff feels that the two personnel doors at the main operating level are in the category of closcres which it is desirable to have function. The control circuitry has therefore been made automatic and to a degree redundant.

I These doors, which are the only doors which can be opened and remain open unattended, are the doors through which evacuation of the building would be expected. Manual back-up provisions exist to enable operator personnel to close I- these doors if automatic cle'sure fails. Door position is indicated at control stations to provide further notification of malfunction at the time of an accident. Sir.cc they can be routinely tested for operation the likelihood of failure can be measured and maintained at a low level. In any case, however, both of these doors were designed to provide positive closure of arcaway vestibules which are themselves a double 8 door arrangement meant to restrict air flow under normal operation between the confinement building and the laboratory wings. This provision was made in the design to effect air control to limit spread of contamination from areas of relatively more probable radioactive spill to more likely clean areas.

Tests were performed with the vestibule doors closed and the sliding door open to determine the pumping speed margin in a manner similar to the previous tests of building closure failures. The emergency exhaust system was operated auto-I matically to mr'sure exhaust rates and internal building pressures. Building pressure stabilized at approximately

-0.12 inches with an average of 165 cfm exhaust rate for the E north-east entrance (1205/2 door) and approximately -0.13 g inches at an average of 140 cfm exhaust rate for the south-east entrance (1212/2 door). Since these numbers are com-parable to the results for the previously tested closure f,ou I the reactar confin wat systtr is capable of 'op t operation witho -isk of un ontrelled leak w .

do n ll I o_.,

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(

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I The NBSR staff feels therefore that only very little would ,

be gained if any by keeping these doors routinely closed.

1 The frequency with which they must be opened negates their function as closures and it is at least questionable that constant use enhances their likelihood of closing when desired.

Since it is not absolutely necessary that these doors close to maintain control of an emergency situation as it might be in the case of a containment building that does not have provision I for concrol}ing the leak of a building, the NBSR staff feels the original design concept for the NBSR is proper and adequate.

Additional indicating lights have been installed on the emer-g gency fan panel for the five reactor building doors. These g lights inform the opera. tor of door seal status. Pressure switch R-1 in the door seal control schematic shown in Figure 6.1 gives this remote indication. No alarms have been provided in either the control room or the emergency fan panel, since three of there doors are equipped with automatic reclosing mechanisms, if opened, during a major scram condition (see I Figure 6.2). The truck door can not be opened during a major scram or a Reactor "on" condition (see Figure 6.3). The emer-gency personnel door is also equipped with an automatic re-closure mechanism (see Figure 6.4).

The thermal overloads for all but the emergency personnel exit door have been removed, to prevent trips from this source during closing conditions.

2) Valve ACV-10 which provides EF-5 and 6 suction from the process roem has been revised to give remote manual operation cf this valve from the emergency fan panel. One would expect the highest activity to be in the process room. Mixing and clean-I up of this activity normally would be accomplished by the internal recirculation fan SF-19. This fan would also equalize pressure throughout the building. During these conditions ACV-10 would be left closed, giving EF-3 and 6 suction from lesser activity areas.

3) Figure 6.5 shows the added DSR-1 contact in EF-2 dilution fan I electrical schcmatic. This contact will open to shut EF-2 "off" on a major scram signal, thus preventing any exhaust from the reactor building resulting from an ACV having failed to close. The operator is informed of the status of these ACV's from the emergency fan control station. If all are closed, EF-2 can be restarted from the emergency fan panel to provide additional dilution of exhausted air.

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I QUESTION 7(a)

Discuss the additional personnel and engineering support services I to be available for initial startup and operation. Provide an organizational chart showing the relationships of these personnel and services to each other and to management. Discuss the adequacy of this startup and initial power organization.

. RESPONSE In NBSR 9, Final Safety Analysis Report on the National Bureau of Standards Reactor, certain services to the Reactor Operations Section were briefly outlined and tabulated.

I These may be found on pp 1-4 and 1-5 under " Organization." It is the purpose of this response to elaborate upon these services I and further to identify others. First we present the most current organizational chart of the National Bureau of Standards.

This is shown in Figure 7.1, where again the organizational units important to Reactor Operations are underscored.

I 7(a).1 REACTOR RADIATIONS DIVISION Within the Division which manages the Reactor Operations I Section are other sections or units which contribute to the safe and proper functioning of the reactor. These are:

7(a).l.1 SCIENTIFIC SUPPORT. This Section establishes and applies methods for the measurement of radiation fields within the reactor. It may also utilize the developments of other Sections within the Division, e.g., those of the Neutron Nuclear Physics Section. To date the Division has developed two new methods for

}

the measurement of neutron flux. One of these is for the measurement

,j of thermal flux, and the other is for the measurement of fast flux.

i3 In addition to applying these new methods, the Section will apply the older methods of foil and wire activation.

The Section also effects the development and installation of new devices and instruments for the reactor. It is the responsibility of this Section to satisfy the need for any new I or special instrument which for any reason may transcend the capabilities of Engineering Services. Normally this would only be in the case of an untried instrument based on an unexploited phenomenon.

7(a).1.2 ENGINEERING SERVICES. This Section provides the principal technical support to Reactor Operations. Not only I does Reactor Operations have first call on the mechanical and electronic service capability of this Section, but the Section assumes prime responsibility for the basic mechanical-electrical 7-1

J I integrity of the plant. It is this Section, for example, which checks out, calibrates, etc. all of the nuclear and process instrumentation; moreover, many of the vital mechanical components of the reactor system were designed, fabricated, installed and tested by this Section. These components included, for example, the grid plates, the fuel handling tools, and the fuel transfer or dropout system.

In addition to providing this design and hardware service, the Engineering Services Section trains key reactor supervisors in the proper method of test and manipulation (e.g., nuclear instrumentation, fuel handling, etc.).

7(a).1.3 HAZARDS EVALUATION COMMITTEE. This committee is composed of knowledgeable senior scientists and engineers of I the Division as well as the Senior Health Physicist assigned to the NBSR. Its membership provides experts in nuclear engineering, I mechantcal engine'ering, electronic instrumentation, nuclear physics, and health physics. The committee is~ chaired by the Deputy Chief of the Reactor Radiations Division and will meet at least once per month to consider proposed new experiments and/or changes in the e

r I plant. Detailed records (e.g., each members opinion) are kept of each action or recommendation to the Division Chief. Thus no new experiment or change in the reactor plant is effected without I the approval of the Division Chief who in turn is advised by the committee.

g This same Committee will also make the final determination

g for the Division that the reactor plant is operational initially.

Its detailed composition is discussed further in 7(b), below.

7(a).2 NBS DIVISIONS OR UNITS 7(a).2.1 0FFICE OF RADIATION SAFETY. This central office I administers all health physics services at NBS and evaluates the.various radiation manuals generated by the Chief Health Physicist and his staff. The Office is governed by a Chairman appointed by the NBS Associate Director for Technical Support, I and a committee whose membership is currently composed of the Chief Health Physicist and the following division chiefs:

Analytical Chemistry,- Reactor Radiations, Radiation Physics, Plant,'and Administrative Services.

Health Physics personnel are permanently assigned to the NBSR I by':he NBS Chief Health Physicist. This NBSR group is staffed for 24 hr. operation and is headed by the Senior Health Physicist, a member of the aforementioned Hazards Evaluation Committee.

B The Health Physics Group performs all of the usual health physic, operations as well as the monitoring of the building zonc s, I wet waste system, outside air and ground water, etc. It also affects the required relationships with nearby hospitals for the care of injured personnel who have been either irradiated and/or I contaminated. It has independent authority to require corrective 7-2

+ .

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action for any situation which is in' violation of the Bureau's l

E Part;20 License.

l

!7(a).2.2 PLANT DIVISION. The Plant Division is capable I of making requested alterations in the non-nucI' ear aspects of the NBSR; i.e., this is a technical division with general mechanical and electrical design and maintenance capability. Unlike most of the NBS buildings, however, the NBSR cannot be altered or adjusted (e.g., air conditioning) without the specific approval - " -

of the Reactor Operations Chief or the Reactor Radiations Chief.

Close cooperation exists between Reactor Operations and the NBS Plant Division. Experience has proven the ability of the Plant Division to provide even certain nuclear related I services. For example, the Plant Division has undertaken the job of procuring new absolute filters when needed, including theirl shipment first to Oak Ridge for testing.

7(a).2.3 ADMINISTRATIVE SERVICES DIVISION. The Chief of

. this Division was identified above as a member of the committee I which evaluat;es the functions of the Office of Radiation Safety.

The Division is responsible for security and safety exclusive of radiation safety. It provides guard, fire, and ambulance service and effects the needed relationships with the Montgomery I County Fire Board and Police.

7(a).2.4 ANALYTICAL CHEMISTRY DIVISION. This Division has i I extensive capability in all types of analysis and provides such service to the Bureau. In particular the Division will provide on a routine-and demand basis the off-line analysis which the I Reactor Operations Section requires of both the primary and secondary systems. This includes the mass analysis of the heavy water as well as all of the analysis required for the identification of corrosion and gas products which may be found by chemical monitor-I ing of the NBSR water and gas sy0tems.

7(a).2.5 INSTRUMENT SHOPS DIVISION. Shops constitute a central Bureau Service which operates a large instrument shop and i various smaller area shops. The larger reactor. mechanical jobs are fabricated in the central shop, and the smaller and more g immediate jobs are performed in the NBSR area shops. The latter are of two types: cold and hot. Each of these shops can accommodate

$ two machinists who are assigned to the Division by the Instrument Shops Division. As indicated above the services of these men are controlled by the Chief of the Engineering Services Section, whose first responsibility is to provide support to Reactor Operations.

7(a).2.6 OTHER NBS DIVISIONS. NBS possesses broad physical, chemical, and engineering capability. From time to time the expertness of NBS has been required ,on a non-routine basis and~

was always found to be immediately t'orthcoming and available

-I to the Reactor Radiations Division. Some examples follow.

7-3

o .

I 7(a).2.6.1 Metallurev Division. Analyzed and recommended 1 the proper material and the method for installation of sacrificial anodic spools for the various heat exchangers in the process systems.

.I l 7(a).2.6.2 Building Research Division. Sampled, tested, and made recommendations for the repair and addition of certain paint I coatings within the reactor building (see response to Question 3).

7(a).2.6.3 Mechanics Division. Analyzed and provided data E on the mixing of air inside the reactor building, g

7(a).2.6.4 Computer Services Division. Provided computer

. service in the digital calculation of neutron fluxes within the

! reactor core and in the analogue calculation of core dynamics.

7(a).3 OUTSIDE SERVICES 7(a).3.1 THE NALCO CHEMICAL CO. The Division holds a con-tract with the NALCO Chemical Co. for materials, consultation, and analysis in the treatment of the secondary water system.

This company has been in business since 1930 as a large chemical supplier. The Industrial Division has its headquarters in Chicago with offices and representation in 250 cities throughout the country.

Its international activities are equally extensive. Among other things the Division supplies chemicals for the treatment of cooling I water systems. Along with these materials this division of the company provides consultation and analysis on the systems it services.

I 7(a).3 2 SAFETY REVIEW COMMITTEE. The Division engages the services of three outside senior consultants for initial and periodic review and evaluation of the Reactor Operations Section and also of actions taken by the Hazards Evaluation Committee.

Currently these men are:

Mr. Robert Powell, Chief, Reactor Operations, Brookhaven National Laboratory.

Mr. Frederick Martens, formerly Chief, Reactor Operations, Argonne National Laboratory.

Dr. William McCorkle, Director, Radiation Laboratory, Ames National Laboratory.

I Personnel of comparable quality will be sustained on this enmmittee, Minutes of thic ccanittce's meetings and their re-commendations will be kept for review by the Division Chief and others.

The cotmittee will first meet at the NBSR prior to fuel loadinc and af ter the Hazards Evaluat ion Comnittee has approved the init c. .

of startup. In this way the Division Chief will be further appraised I

7-4

?b o of the feasibility of startup prior to ordering same. The committee will aslo be requested to review and approve the zero power results and any other results which the Division Chief wishes to be reviewed during the plants assent to power.

Shortly after steady full power operation is achieved the I committee will be requested to perform another review of operations.

Thereafter the committee will meet at least once per year at the NBSR for general,and thorough review of all phases of the operation including Health Physics and the recommendation of the Hazards Evaluation Committee.

7(a).3.3 OPERATIONS CONSULTANT. ;n the initial period of I startup (i.e., criticality, assent to power, and initial operation),

the Division will engage the services of a senior consultant to act as Overriding Supervisor for the Re; actor Operations Section.

I This consultant will have proven experience in startup, operations, and management;of course, he shall be acceptable to the AEC Division of the Reactor Licensing.

Proven experience constitutes previous successful super-vision of startup and initial operation of a reactor plant or system of comparable complexity to the NBSR. This means that in addition I to having been the on-line responsible individual for the actual startup, it was also this same individual who approved all of the operating and other manuals as well as the physics and other tests necessary to, prove the reliability and safety of the plant. At I the NBSR he shall again assume this total responsibility, being responsible only to the Division Chief.

7(a).4 STARTUP AND INITIAL POWER ORGANIZATION All' of the above cited forces are brought into play during I the initial period of criticality, assent to power, and initial power operation. At the termination of the light water systems tests and at the time of system drying, refilling with heavy .

water, etc. (all of which is just prior to fuel loading) the I internal Hazards Evaluation Committee, the outside Safety Review Committee and the Operations Consultant will have given their approval to the Division Chief for startup. This means that the entire system, staff, manuals of procedure, etc. will have been I reviewed and found acceptable.

I The outside Operations Consultant will then direct the fuel loading, zero power tests, etc. as outlined in the Startup Procedures. Data is analyzed by the Hazards Evaluation Committee and reported to both the Operations Consultant and the Division Chief. Members of the Hazards Evaluation Committee will also participate in the taking of nuclear and other data. Personnel from the Sc ient if ic Support Section will be available to serve I in this capacity also as directed by the Division Chief and the Operations Consultant. Continuous service from other parts of 7-5

a ,

In the Bureau will be available as indicated in 7(a).2 above.

particular, analysis of the isotopic ratio in the heavy water as well as frequent chemical analysis of the water and gas systems I will be performed.

At the conclusion of the zero power tests or at any stage i

in the assent to power which the Division Chief or the Operations I Consultant deems is in need of review by the Safety Review Committee it will be so required. A chart displaying the lineBroken responsibilities during this period is given inItFigure should 7.2.

be noted lines represent of f-line available services.

a that for this period Engineering Services is placed in-lineTheseunder the Operations Consultant along with Reactor Operations.

I various forces and the chain of command is deemed adequate for this period. .

QUESTION 7(b)

Describe the composition of the Reactor Safety Committee

• its charter and organizatinal po s ition, and the criteria established  ;

I to insure that a portion of the membership will be independent of any responsibility for operation, use, or management of the e

RESPONSE

The Hazards Evaluation Committee is currently composed of I the following personnel.

Chief, Reactor Operations I Chief, Engineering Services Nuclear Instrumentation Engineer, Engineering Services Chief, Neutron Nuclear Physics Chief, Neutron Solid State Physics I Senior Health Physicist, NBSR The committee is required to review all new experiments or pro-I posed changes in the reactor plant as cited in 7(a) .1.3. Whether or not changes have been proposed the committee will meet at least once per month to review operations generally and to examine the I log and other records of reactor operations. Recommendations to the Division Chief could result from such general review.

Of the six personnel cited above the two engineers from the I Engineering Services Section and the Senior Health Physicist are independent of operation, use, and management. Their function is solely to produce equipment and conditions which are both safe and proper. Furthermore, should an experimental proposal or change I be placed beforc this Committee by one of its scientific user members, that member is automatically disqualified from the delib-I erations of the Committee on that proposal.

Misnamed by DRL--should read Hazards Evaluation Committee.

7-6

I Finally, it should be recalled that the actions of the internal Hazards Evaluation Committees are periodically reviewed by the outside Safety Review Committee. This will provide the Division Chief,with advice as to the soundness of the technical judgements I made by the Hazards Evaluation Committee.

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I QUESTION 8.

Your October 1,1966, reply to our previous question regarding the maximum reactivity insertion whose energy could be contained without rupturing the reactor vessel or primary system should be I expanded to include transient as well as steady-state considera-tions. If, as it appears, the vessel will be contained from gross strain failure by the backup biological shield, the re-g straining effect of this shield should be considered. Similarly, 5 failure of the beam tube nozzles should be considered, and the effect of their failure should be factored into the analysis.

Also, consider the effects of vessel or top plug jump. ,

I The response of the NBSR primary system to the transient phenomena associated with a sudden reactivity insertion has been examined both by the technique of explosive equivalents as developed I by Wis data.

p b roctor et al (1) and by correlation with the Spert II It is noted that the direct relationship between a TNT explosion and a nucicar excursion' is rather oblique. The conservatism which is a necessary and laudabic goal of the explosive equivalent I techniques borders on the excessive when applied to the excursion energy profile of a D 02moderated reactor of the NBSR type.

I This approach then represents an ultraconservative approach to the problem. The only directly applicable excursion data is from the Spert II (1) series of tests. portions of this data which I characterize the pressure pulse and water hammer ef fect are un-fortunately sparse. This data does however amplify the basic fact that on a Mw-Sec for Mw-Sec basis, D 0 moderated and H 2O 2

moderated reactor excursions are of distinctly different character.

I - The peak pressures measured during the Spert series were less than those predicted by explosive equivalent techniques (and correlated to light water accidents) by factors approaching 10.

Although the NBSR vessel and top plug assemblies are quite complex from the standpoint of configuration and penetration, the limiting features of the system are quite decernabic. The I wide margin of energy necessary to breech the primary system in comparison to the energy necessary to cause top plug movement makes it quite clear that plug movement is the limiting feature.

Failure of the vessel proper is limited by the restrain-ing effect of the thennal shield and the thermal column D 20 I tank. These items are separated from the vessel wall by approx-inately one inch. Use of the Explosive Containment Laws for

(

W. R. Wise. J. F. proctor et al. U. S. Naval Ordnance Laboratory Test Reports62-207 and 63-140

(

V. W. Goldsbe rry, IDO 16990 8-1

k Nucicar Reactor Vessels developed by, Wise, Proctor et al shows that the vessel itself is capable of withstanding excursions from e 40 to 60 Mw Sec. The strains occurring from much lower energy

[ releases are more than adequate to cause the vessel to reach the restraining surfaces. Once restrained, the energies required to y

fail the vessel are orders of magnitude greater, l

Vessel jump is virtually prohibited since the upper plug system is attached not to the vessel but to the shim ring which -

is in turn attached to the thermal shield. The vessel flange m is clamped between the plugs and the shim ring, but the load capability of the bolts (24-1" dia.) attaching the shim ring

- to th'e thermal shield plus the very high energizes necessary L to sever all the vesset eigtna dictate the fatture of plus bolts (24-1" dia.) long before the vessel could move, a

Beam tubes, including the cryogenic tube and grazing tubes were examined from the standpoint of collapse due to external pressures, e.g., internal vessel pressuras. In all cases pres-F sures over 1000 psi would be required to cause failure.

L Lower vessel piping, subpile room piping and all primary system piping was examined from the standpoint of bursting pressure. For the worst case, e.g., thinnest wall and largest diameter the calculated bursting pressure of this piping is in excess of 1000 psi. For the limiting case for the system, e.g.,

top plug jump, there would be some straining of the piping but no breech of the system.

The interrelation of plug frontal areas, hold down mechanisms, and plug positions is such that, with some noteable exceptions, the limiting feature is the movement of the entire upper plug system. The exceptions are those plugs which were unrestrained at the time of the analysis. Modifications of these plugs and the

( surfaces on which they mount are now underway to provide these plug assemblies with the capability of realizing the full potential f of the rest of the system. This is a very simple task due largely L to the relative smallness of the plugs in question. Flanges will be welded to the plugs to accept the hold down screws and the main plug surfaces will be drilled and tapped to equal the screw strength.

Examination of the possible modes of explusion of the upper plug assembly led to the following most sensitive case, e.g.

least peak pressure required to expel the plug assembly to a given height.

No credit is taken for the 24-1" dia bolts which attach the

{ icwor outer plug to the thermal shield.

An isotropic expanslan of the blast gasea is considereJ t4 hold for the e':plusion period.

L W. R. Wise, J. F. Proctor et al. U. S. Naval Ordnance Laboratory Test Reports62-207 and 63-140.

8-2

o e i

A peak pressure is then calculated which would be nece9sary to eject the entire upper plug assembly to a height of 30 feet (the point at which the plug would contact the roof of the con-finement building). Ttis pressure is then used as the lower I limit of *he pressure which any smaller plug must tolerate be-fore becoming a secondary missile. The energy necessary to eject the entire plug assembly to 30 feet is calculated as approximately equal to the 400 psi, aquivalent of 40 pounds of TNT or approxi-I mately 80 Mw-Sec.

.The TNT equivalent energy release necessary to move the I NBSR plug system is then a measure of the gross upper limit of the significance of an excursion.

I Calculated plug movements for various TNT energy equivalents are given below:

I Mw-Sec 10 20 APPROXDIATE TOTAL PLUG MOVEMENT 0

0 I 30 40 70 0

7 ft. (Plug bottom at floor level) 22 ft.

80 30 ft. (Plug top just below roof level)

I The main upper plug structure of the NBSR is shown in Figure 4.1 of NBSR 9. The weights and sizes of the various plugs subject to explusion and the restraining mechanisms at the time I of the analysis are tabulated on page 8-4. The plugs which are listed as unrestrained are now being equipped with restraints to I assure the retention of these plugs to a minimum vessel pressure of 500 psi, e.g., to match the minimum holding power of the other accessory plugs. These additions are shown as Added Restraints.

In addition, the 4 inch thick floor plate for which no credit has been taken, will be fastened to the main upper plug to provide an effective missile barrier should any of the accessory I plug hold-dams fail. The floor plate plugs will then be equipped with 1/2 inch shear pins to hold them in place.

1 I

8-3

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M M M M M M M M M M e m M M M M e W .

WEIGHT ITDt NO. SIZE KIPS RESTRAINT ADDED RESTRAINT Lower Shield Plug 1 8' OD x 4h ID x 15 24-1"-8 bolts Upper Shield Plt:g 1 8' OD x 4' ID x 40 None Inner Shield Plug 1 52" OD at bottom 21 12 -- 5/8 -- 11 Lower Shim Arm Access Plugs 2 22" x 10" x 30" deep 1.2 2-3" U at 6 II/ f t x 26" Lg for each U r Shim Arm Access 2 22" x 10" x 32" deep 1.3 Lower and Upper 14 .

Experimental Plugs 2 ea 4.25 OD x 30" Ig. . 1 None 1/2" Shear Pins position 3[" Exp. Pos. Plugs 7 3[OD x 5' Lg. . 225 None 3-h" Bolts 2[" Exp. Pos. Plugs 4 2h OD x 5' Lg. . 125 3- -20 bolts Fuel Element lland Tools 24 1 OD x 4h' Lg. . 050 3-k-20 bolts Fuel Transfer Plug 1 6" OD x 7' Lg. . 5 None 3-h" Bolts P

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A

5 l QUESTION 9.

Your October 1, 1966, reply to our previous question regarding 8 the failure of a shim arm should be expar.ded to include con-sideration of the consequences of the maximum reactivity tran-sient which would occur under any rod programming scheme. These I consequences should be compared with the energy absorbing cap-ability to contain or limit the consequences of the excursion.

If, as you indic*ated in your previous reply, it is necessary to provide means to assure reactivities do not exceed acceptable I values you should include your analysis of more positive means than administrative control. For example, consideration could be given to combinations of such schemes as:

(a) rod programming interlocks; (b) rod steps which prevent a broken rod's falling; (c) negative period scrams; (d) reduced control rod worths; I (e) any others.

provide complete details on any schemes employed.

In responding to the question of shim arm failure in the report NBSR 9A credence was given to such an occurrence for the sake of examining the margin of safety for the NBSR. It is I actually very difficult to imagine a credible circumstance for such failure, when the actual arm configuration is examined. To provide additional assurrance, however, " rod stops" have been I designed to be located on the lower grid plate to restrict rod motion in the event of rod failure. The design of these " stops" is discussed at the.cnd of this question. The scope of such a rod f ailure accident is reviewed below, however, to provide I understanding as to the nature of severe transients in the NBSR regardless of the credibility of such an event.

I Upon examination it is clear that two distinct conditions would exist under which a severe excursion could be imagined.

The most serious situation would involve an operating core in which sufficient fission product inventory has been accumulated I to cause a core meltdown should a loss of coolant occur because of an accident. The e.sact moment in the core operating cycle when such a condition first becomes possible is difficult to ascertain since it involves the question as to just how coolant is lost and how much decay heat is dissipated during a toolant systm f a ilure. It :m s rcasonable to assume that a potential I meltdown condition would first occur scmetime af ter xenon poison I

9-l

I has reached equilibrium, i.e., after approximately twelve hours I of operation.* Under these conditions it could be imagined that one shim arm is holding all availabic excess reactivity (although this violates normal operating procedures). The excess reactivity I would be approximately 4.1*/. as determined from Table 4.6-6 on page 4-17 of NBSR 9. The entries of this table which would apply are the U235 burn-up for the cycle, the reactivity for control at the end of the cycle, and the portion of the reactivity which I is available for removabic expedment poison.

As discussed in Question 1 of N3SR 9A a shim arm fracture I under this condition, if it were free to fall completely through the core, would result in a decrease in reactor power initially (i.e., a negative period) followed by a rapid rise in power. J3 I It is difficult to calculate the exact transient with the ramp insertion rate that would exist with a shim arm falling at its terminal velocity in the reactor moderator. The ultimate period would be approximately 22 milliseconds if no reactor power icvel I or period scram occurred. It is not possible to analyze this transient properly since Spert II data does not exist i.n this range of period. A reasonabic extrapolation of Spart II data I corrected for NDSR core parameters can be made however using the curves of page 129 of the report 100-16990 previously referenced. The total energy release would be approximately I 55 W-sec. As discussed in Question 8 above it is not expected that such an excursion would damage the reactor vessel although top plug movement would probably exist. The energy of the ex-cusion is 60 per cent more than required to melt the core if no energy is lost to the coolant but an appreciabic fraction of the energy would ese ne and only partial melting would be expected unless coolant is iost subsequently. This accident, therefore, I although severe would not be different in its total cmsequences to the off-site public than the already assumed slow core melt-down.

The second condition that it is necessary to consider would involve an accident similar to a startup accident in which a greater reactivity insertion is possibic because the core has not yet developed a full complement of fission product poison.

I- If it is postulated that a shim arm fails soon after startup before temperature and xenon poison has built up then a maximum I reactivity increase would be expected. The magnitude of the reactivity inserted might be approximately 9 or 10 per cent de-pending on experiment loading. The period which wc;uld result I from such an excursion would be approximately 9 or 10 milliseconds.

The energy release would be of the order of 70 W-sec t. sing the same procedures as above. As discussed in Question 8 rhis energy release is less than that required by very conservative estimate I This assur e s a startup h is atturred af ter a norral s!'u t d men I period during shich three fresh elements have been adde <! to the core. The core has been shutdown for at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to allow xenon poison to decay before re-start.

. . 9-2

I to lift the central plug to the confinement building ceiling.

Structural damage to the confinement building is con equently unlikely. The energy in the excursion would be dissipated quickly to the coolant and even though partial melting would I undoubtedly take place in the core, a total meltdown from decay heat is unlikely because a long lived fission product inventory would not exist. Consequently even though this excursion would I be even more severe the danger to the off-site public is still no greater than previously discussed. -

ROD STOP DESIGN In the previous design a broken segment of the NBSR shim safety arm would have been free to move down and along its pro-I jected length to the lower grid plate. The weight of the broken segment and the lack of any significant coolant flow in the area account for the downward movement. The fact that the shim I safety arms (1 inch wide x 5-1/2 inches high x 69-1/4 inches long) operate in a space 3 inches wide which is bounded by fuel elements on both sides for the length of the arm make any other mode of movement impossibic.

By extending the existing shim arm guides down to the lower grid plates and then back up to the adjacent guides (Figure 9.1),

I any broken portion of a shin safety arm can be trapped in its most effective position. A minimum width of 1-1/2 inches will be nalntained for the guide extensions (Figure 9.2), thus pre-I venting any broken segment from slipping down past the guido extension surface. The 3/4 inch cicarance between the guide extensions and the fuel elements will assure retention of the 1 inch wide shim safety arms.

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QUESTION 10.

Provide an analysis of the tritium doses resulting from a massive heat exchanger failure in which the heavy water carries over to the secondary cooling tower.

! RESPONSE The nature of the problem associated with primary-to-secondary system leakage has been reexamined for the NBSR. The original design of the primary system specified that the primary system pressure should be always greater than that of the secondary system to assure leakage from the primary to the second2ry rather

• than vice-versa. The purpose of this design is to protect the isotopic purity of the primary heavy water coolant. To minimize the possibility of leakage between the two systems, the interfaces of the systems, which occur only at the heat exchangers llE-1 and llE-3, were specifled to be double walled, i.e., the tube sheets of both exchangers are doubic walled. The space between tube sheets in both 2xchangers is examined for moisture by electrical circuit leak detectors which can detect droplets of water and alarm at the control console annunciator.

Af ter fabrication an1 installation the heat exchangers were examined for leakage by mass spectrographic helium leak detector techniques. Initial integrity is, thereby, assured.

During reactor operation the primary coolant becomes radioactive from at least three causes, namely3 neutron capture in the coolant deuterium to produce tritium (11 ), neutron activationofoxygentoproguceradioactivenitrogen(N16) and through accumulation of Na 2 following a neutron reaction in aluminum. Of these three isotores only tritium is suf ficiently long lived to be a potential radiation hazard. The fact of the f production of these activities permits a back-up system of Icak L detection, however.

Calculations show that the concen . rations of the above three isotopes in the primary coolant at the exit of !!E-1 are as given in the table.

Table 10.1 1sotope Conc.

N 1.2 c/cm 2* "

Na 0.0 12 11 4.2 x to (at saturition after > 20 yetra) i 10-1

I These calculations are for 10 FN operation at saturatic of I each activity allowing gr decay in pipe transit from the r  ; tor vessel in the case of N . As indicated in previous submittals the hazard to the public associated with spillage of primary coolant are minimal with these activity levels. Off site I concentrations of tritiated heavy water vapor resulting from saturation of process roon air and subsequent release are a small fraction of MPC levels of 10 CFR 20. With these considerations j

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I in mind it is to be expected, as confirmed in calculations presented below, that any release to the secondary system will not be hazardous since dilution of the stated concentrations is to be expected by the secondary water.

As stated in NBSR 9, a radiation monitor, RD-3-1, is located on the secondary side of the cooling system to detect leakage of I heavy, water to the secondary side. A calculation has been made of the leak rate to the secondary which the monitor can reasonably be expected to detect. The calculation assumes the concentration I of N IO stated in the table above to enter the secondary and mix with the '.950 gpm flow expected {or 10 FM operation. Assuming a minimum sensitivity of 3 x 10" c/cm3 for the detector (which level has been measured during cattbration with Na22, a considerably lower energy beta decay emitter than N16) the maximum leak rate to the secondtry would be approximately 1.3 cm 3/sec. At this rate only a small f raction of a curie would be relcued to the secondary I system before corrective measures to isolate the heat exchanger and reduce secondary flow to the shutdown condition could be initiated.

I N 16 I the leak shoald develop while the reactor is shutdown then detection is no longer g ef fective technique.

shutdosn flow conditions Na detection becomes a practical Under the scheme, howeverS and the maximum Icak rate that could occur would be 1.2 cm /sec.

Since Na ' has a half life of 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, leak detection I sensitivity would decrease during shatdown conditions.

sampling of secondary water can introdu:e a very powerful technique for indicating the existence of incipient leaks, however.

Periodic For example, tritium counting techniques of standard volume of 3 l water can disclose tritium concentrations as low as 10' c/cm ,

l a factor of 30 below the MPC of 10 CFR 20. When this sensitivity

! rimary system in the shutdown condition, a leak of 1.9 x 10-6iscmused to calculate a leak rate from the g/sec can b found.

This feak rate is calculated at the beginning of a leak and no credit is taken for buildup of tritium in the secondary which I would lower the minimum detectable leak rate considerably depending on assumptions as to the conditions under which it would tako place.

This leak rate is already in the category of a weeping leak of very small magnitude and would indicate potential heat exchan.:er I fatlure at an eirly t,,.

In any event it , n be eA 2d sh it tre the con,e p n  ;

pe rn i t t i n g a .ak it jo it the ninimo detectable le"e l a i measured by N h activity to go unnoticed for a time equal to that at which the Na24 level in the seconda ry becomes de*.cc tab le.

I 10-2

I This time would be approxinately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The tritium concentration I in the secondary would build up to 0.6 c/cm 3, approximately 200 MPC for unrestricted areas.

I The tritium concentration in the air at the top of the cooling tower flue can be calculated assuming the air to have initially a relative humidity of 507. and to become saturated with secondary system tritiated water. An air temperature of 90 F would result I in a concentration in air of 9.5 x 10-fbc/cm3 . This concentration is less than twi,ce the restricced zone MPC level. Atmospheric dilution of the cooling tower plume would quickly reduce this cancentration as it moves from the tower.

One can examine the meaning of such icvels of contamination I of the secondary water by calculating body burden and total dose for a person inadvertently drinking withour further dilution at a normal daily rate for the period during which the leak goes undetected. The body burden that would result would be 270 c and the total dose 0.054 rem.

Finally the question of the consequences of a sudden large I leak to the secondary system can be examined. If one considers the rupture of a singic tube in the large heat exchanger, llE-1 at the point where it penetrates the back tube sheet, primary coolant I would be released through both ends of the ruptured tube at approxi-mately 55 gpm. This flow caused by the higher primary system pressure would continue until the primary pumps were turned off.

The following actions would occur automatically (a) the " Reactor Vessel Overflow Loa" annunciator would sound immediate ly, (b) the "}!i Radiation-Secondary System" annunciator would sound almost immediately assuming equilibrium tritium 3

concentration, 4.23 mci /cm , in the primary system, (c) 4.2 minutes later the " Reactor Vessel Low Level" annunciator would sound, I (d) 8 2 minutes af ter the rupture occurred a reactor

" rundown" would be initiated, and (e) 16.9 minutes af ter the rupture a reactor scram and I primary pump tr!p "off" would be initiated. One primary shut down pump would come on autcmatically.

I A total of 831 gallons of D 02 would be discharged to the secondary within this elapsed time. If the main secondary pumps were not tripped of f by the operator prior to the primary pumps trip, the leak would reverse to degrade the D2 0 with light water.

Assuming this did not occur and the n iin secondary punps wre tripped eith the mond try shutdan pump in ope ration, the D0 I

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I leak rate would be reduced to 25.5 gpm until llE-1 was isolated, a time which is somewhat uncertain but can be estimated to be I an additional 20 minutes. The operator must open the process room door and isolate the secondary side of IIE-1 by closing valves SCV-117 and SCV-120 (See Figure 6.1 of NBSR-9). During I this time an additional 510 gallons of D 20 would enter the secondary, for a total of 1341 gallons ot concentration of 4.23 mcf/cm3 .

D2 0 at a tritium I The tritium concentration off-site af ter the normaj atmospheric dilution of the order of 1000 would be 1.78 x 10-6pc/cm or approximately 9 times MpC for unrestricted areas. With this I concentration, 250 Fours of exposure would be required to assimilate 1 millicurie of tritium, resulting in a total dose of 0.20 rem.

A break in the purification heat exchanger llE-2 would be less '

severe since the total amount of D20 available to be discharged would be limited to that in the sump of the storage tank which is approximately 326 gallons (See page 5-7, NBSR-9).

The heat exchanger leak detector is presently a single unit without a redundant backup. Normal operation will be indicated I by the level of background radiation. Since nomal operating procedures call for it to be inspected for operation each shif t warning of detector failure would be expected. In addition the detector will be checked for functional perfomance periodically using a standard check source.

The NBSR staf f believes, therefore, since the consequences of I a massive Icak are sufficiently limited and the likelihood is maintained low by proper inspection, good water chemistry control and frequent highly sensitive water sampling procedures that the I NBSR design is adequate. 71 the event detector failure is noted more f requent sampling can serve as emergency procedures during the time required for repair or replacement.

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I QUESTION 11.

I In your answer to our previous questions, you indicated that in situ tests of the reactor safety system at the maximum control room temperatures you would expect to operate were not planned.

I Although factory tests were conducted it is our opinion that elevated temperature tests should be repeated to detect possible cabic faults, cold solder joints, dif ferences in compor.ent cool-I ing arrangements, proximity of heat producing equipment, etc.

Please inform us in detail of your reasons for not performing such tests if this is still your intent. ,

RESPONSE

I

1.1 DESCRIPTION

OF TESTS Thermccouples were installed at 16 points in the console to monitor temperatures of the reactor ( fety system at selected points. These points were as follows:

POINT LOCATION 1 A Left Bottom Botto.n Nuclear Cabinet 2 A Right Bottom Bottem Nuclear Cabinet 3 A Left Middle Picoammete r chassis 4 A Right Middle Cabinet Metal Temp.

-10 V Power Supply I

5 A Left Top 6 A Right Top F10 V Power Supply 7 A Top Air Ambient Air Temp.

8 A Middle Air ambient Air Temp.

9 E ECF Chan FRC-3 Monitor SW 10 E ECE Chan FRC-4 Monitor SW 11 J Top Cabinet Ambient Air Temp.

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12 13 E ECH E ECA Chan LRC-1 Monitor SW Chan TIA-4'6' Monitor SW I C CCD 14 Chan FRC-1 Monitor SW 15 D AN4 Scram Annunicator Painel .

16 F 42V Supplies Process Instrument Power (1)

Re f. Fi cu: a 9.1 Nd h a.

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11-1

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The output of the 16 thermocouples were indicated on a 0-10 I MV multipoint recorder. An ice bath cold junction was employed to maintain a constant re ference temperature.

I The entire control roo.n was heated by the installed heat-ing system to approximately 95 F. The installed system was suppicmented by the use of three 1600 watt electrical heaters.

The temperature attained an equilibrium of 107 F af ter 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br />.

I The temperature was maintained at 107 F for a 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> period the rea f te r. -

The trip points of the following devices were checked for drif t and change in calibration as a function of temperature.

DEVICE FUNCTION FRC-1 (FA-1) Reactor Outlet Flow Scram FRC-3 Reactor Outer Plenum Flow Scram FRC-4 Reactor Inner Plenum Flow Scram LRC-1 Primary Coolant Level Scram IIA-40 Reactor Differential Temperature Scram 07-15 NC-3 Log N Amplifier I 07-13 NC-3 Period Amplifier 07-16, NC-4 Log N Amplifier 07-14 NC-4 Period Amplifier 07-39 NC-6 liigh Flux Scram 07-42 NC-7 liigh Flux Scram 07-45 NC-8 High Flux Scram The trip points of the above devices were checked once per hour for drif t by introducing calibration currents into the various channels. In addition the annunicators were checked for operability.

I During the test the nuc1 car instrumentation -10 voit power supply tripped out caused f rom an over temperature trip device in the pass transistor assembly. This trip occured af ter I approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and caused the rod clutches :o de-energize.

The room temperature at this time was approximately 102 F. The ambient air temperature at the top of the cabinet directly over I the supply was measured by thermocouple No. 7.

at this point was 126 2 F.

The temperature The ambient air temperature at this point normally operates at 96 F with a 72 F room temperature.

The -10 volt supply was reset several times and would stay on until the rbtent tc -pe r itu re of the supply reached 126 F.

the time that the aupply would rem tin on would vary with the I time the supply was de-energized.

. . 11-2

I The maximum change in setpoint .of the high flux scram trip I units was approximately 4 to 6%. At the 1070F room temperature the high flux trip shifted from 130*'. to 1267, of full power which is in the safe direction. The maximum ambient temperature at the high flux trip unit was 113.6 F.

The changes in setpoints of the process instrument scram trips were insignificant. The maximum temperature attained, I 129.7 F, was in the monitor switch for channel TIA-40 (TA-40).

The manufacture lists the maximum operating temperature at 130 F. The trip point drifted from 15.5*F to 15.1'F.

The maximum temperature 139.4 F attained during the tests was at a point directly above AN-4, the scram annunicator panel.

No failures were noted in this equipment.

MAXDIUFf TEMPERATURES Norm. Oper6 Temp. (72 F amb.) Max. Temp 3 (107 F amb.)

7

1. 72.1 102.0

2. 74 110.0

3. 85.1 116.6

4. 68.7 99.8

5. 90 120.0

6. 86.1 104.0 I 7. 96 126.0

8. 80 113.6

9. 70.3 115.0

10. 75 116.6 ,

11. 80.4 117.6

12. 72.8 115.0

13. 86.5 129.7

14. 76.2 105.7 139.4 I 15.

16.

103.98 86.7 122.0 I 11.2 SUIDIARY The temperature tests indicate that while several hot spots are present in the console it is posaible to nonitor the I reactor pir. .i t e r s it t o:t t ro l t p rob 11 11 i t.) of e!>ttinin: to : ,e pee r it u re a up to 102*!'.

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11-3

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The temperature trip out of the -10 V supply produces a jg safe failure and would scram the reactor by de-energizing the clutch power supplies when the control room temperature reaches 4

l 3 l 102 F. Since the clutch power supplies are the greatest load on this power supply in all probability the nuclear instru-I mentation would continue to monitor the reactor with the clutches de-energized at higher ambient temperatures.

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I 11-4

I QUESTION 12 Provide the following information regarding the electrical distribution system:

(a) How long 1an the emergency battery supply emergency loads at its minimum acceptable charge condition? What loads are involved?

I (b) How is switchgear operating power supplied and is this power split such that single failures will not violate the split bus concept?

I 12 (a) HOW LONG CAN THE DIERGENCY BATTERY SUPPLY DIERGENCY LOADS AT ITS MINIMUM ACCEPTABLE CHARGE CONDITION? WHAT LOADS ARE INVOLVED?

The following loads are connected to the 125 VDC supply:

(1) D 0 Shutdown Pump, DP-5 2

(2) D 0 Shutdown Pump, DP-6 .

2 (3) Emergency Exhaust Fan, EF-5 (4) Emergency Exhaust Fan, EF-6 (5) Misec11ancous Power Panel, DCP-2 (a) DC Valve Power (b) DC Relay Power (c) Annunciator Power I (d) DC Instruxent Power Miscellaneous Power Panel, DCP-1 (6)

(a) Switchgear A-Control Power (b) Switchgear B-Control Power (c) Annunciator Power (d) Emergency Lights I (7) Inverter / Diverter The total load assuming only one shutdown pump is operating approximates 106 amps. At the full charge of 2.15 volts /coli the battery in designed to sustain this load for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> resulting in a final voltage of 1.75 volts / cell. The minimum acceptabic charge per cell is approxim.1tely 1.95 volts; therefore, this electrical load could be carried for approximitely 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at th t< mininum charge.

12-1 .

. o I.

I 12 (b) IIOW IS SWITCIIGEAR OPERATING 'PORER SUPPLIED AND IS TIIIS POWER SPLIT SUCil TIIAT SINGLE FAILURES WILL NOT VIOLATE TIIE SPLIT BUS CONCEPT 7 Switchgear operation is manual on all breakers except those feeding MCC A-5, MCC B-6 and the two diesel genera tor feed breakers.

These breakers are spring actuated to close with DC power to re-cock I the actuating mechanism. DC power is fed to MCC A-5 and "A" diesel generator from panel DCP-1 circuit #3 while MCC B-6 and "B" diesel genera tor is fed from panel DCP-1 circuit #5. This panel in turn is fed direcO.y from the 125 volt DC distribution board.

Operating power for the various automatic transfer switches supplying emergency lighting is fed from circuit #1 of panel DCP-1.

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I 12-2

I QUESTION 13.

It is our understanding that all potentially radioactive liquid waste will be batch sampled prior to discharge. Provide sufficient details, including any changes to the system which may be required to allow us to evalua te this approach.

RESPONSE

The hot wa te collection system vill be operated in such a manner I as to divert all suspect liquid waste through the installed radiation de tec tor RD 4-3. Af ter passing through a delay tank, the water will pass through RWV-3 and collect in the 1,000 gallon retention tank.

I From here the waste will be pumped into a 5,000 gallon hold up tank.

Af ter mixing and batch sampling to assure radioactive concentra tions below 10 CFR 20 limits, the contents of the retention tank will be pumped to the sanitary sewer. To achieve this mode of operation two I modifications were affected and are shown on the revised Figure 7.19 of NBSR-9, (1) a blank flange was installed in the line from the delay tank to the sewer and (2) the discharge line of the hold up tank pump I has been reloca ted to enter the sewer just upstream of the limes tone pit.

I The radiation monitor will operate continuously and alarm in the control room on high activity. This monitor also automatically re-routes the effluent to the two ba tching tanks. This re-routing is achieved by automa tic closure of RWV-3 and opening of RWV-6 and 8 to the batching tanks.

Contamina ted liquid was te coliceted in the ba tching tanks can be I recirculated thru an ion exchange column which has been included in the revised sys tem.

The system as installed permitted remote sampling of the contents of all tanks except the delay tank.

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QUESTION 14.

Provide the following information regarding the Emergency Cooling System:

I (a) The muimum and minimum flow rates from the system when the bottom feed method is employed.

(b) A quantitative analysis of the ability of the top feed method to adequately cool the fuel elements.

(c) Modifications required to the system to assist the operator in his determination of whether to use the top or bottom feed method. -

RESPONSE

14(a) THE MAXIMUM AND MINIMUM FLOW RATES FROM THE SYSTEM WHEN ,

THE BOTIOM FEED MEIHOD IS EMPLOYED. s I On May 20, 1966 a drain test was performed to measure the flow rate in the emergency cooling system using the plena feed mode with a reactor level at the height of the upper grid plate.

An average flow rate of 11 gpm was observed. Previous tests on May 19 measured an average of 8 gpm to the outer plenum and 3 gpm to the inner plenum. These values represent the flow rates in the i system throughout the drain time since the reduction in the emer- l I gency cooling storage tank level is small, approximately 4 feet, as compared. to the difference in elevation, 46-1/2 feet, between t the storage tank and the centerline of the reactor core. To con-fim this assumption a test was performed on July 29 and August 1,1966 with no water in the reactor vessel or plena. This was accomplished by opening drains in.both plena. The- test ef fectively increased the head of the emergency storage tank by 11-1/2 feet. A I 12.6 gpm flow rate was observ%, yielding approximately 0.14 gpm per foot of elevation dif ference increase. This information confirms the maximum flow rate was approximately 11.5 gpm with the minimum I set at 10.5 gpm.

14(b) A QUANTITATIVE ANALYSIS OF THE ABILITY OF THE TOP FEED ~

METHOD TO ADEQUATELY COOL THE FUEL ELEMENTS.

The top feed emergency cooling system is designed to back up I the plena feed system in the remote event that the pri. nary system rupture occurred between the check valves in the inlet lines and the reactor vessel. The system and its components are described in Section 7.1 of the " Final Safety Analysis Report on the NBS I Reac to r" (NPSR ). The initial flow o f / 0 gp is distribute uni fo nnly to the 37 core positions. The total flew is niinu c at leist at 25 gro '" ce p le n i s h i a:; the i .ne r rese rve tant -

the emergency cooling tank.

I

. . 14-1

I The most severe cooling problem is presented by one of the 8 elements in the inner ring of the start-up core. At 10 Mw reactor operation this element is running at 571 Kw as compared to 431 Kw for the highest power element in the equilibrium core. (Section 4.6 NBSR 9)

I The total power of the 571 Kw element including all gamma and beta energy is given in the table below as a function of time. The heat removal capacity of the emergency cooling water flowing into each element is I also given. The column headed " Sensible Cooling Rate" is the portion of the available heat removal rate for one element due to heating the emergency cooling wa ter from 106* F to boiling. The other column is the cooling rate available by vaporization of the water.

DECAY HEAT AND EMERGENCY COOLING CAPACITY Time Af ter Total Flow Rate Sensible Vaporization Decay Shutdown Per Element Cooling Rate Cooling Ra te Hea t Ra te gpm Kw Kw Kw 15 sec* 1.08 17.9 164 29 30 sec 1.06 17.6 161 26 17.2 158 23 I 60 see 2 min 5 min 1.04 1.00

.88 16.6 14.6 152 134

-19 15 10 min 68 11.3 103 13 I

• At least 15 seconds is required to drain reactor vessel to top of fuel elements assuming a wide open pipe rupture in the inlet system.

The table shows that the sensibic heat content of the water alone is almost sufficient to cool the element and that the total cooling capacity of the water including vaporization is at least five times I greater than the heating rate of the element. Thus, the cooling capacity ~

is clearly available in the top feed system.

The water is directed into each element by the distribution pan (Section 7.1 and Figure 4.1 of NBSR 9) as a steady stream at the flow rates given in the table. The stream hits one of the heavy' sides of j the element and partially splashes and partially runs down the inside 3 from the point of initial contact (approximately 2 feet above the fuel plates). The 2 foot column formed by the side plates of the fuel element above the fuel plates themselves makes it very unlikely that I steam formed by water splashing on hot fuel plate sections will prevent the water from entering the element. The tendency of the water stream to run down the side pla te also will help to prevent the water flow I from being interfered with by any steam formation.

Although some of the water will splash onto the fuel plates coolin;; them directly, this is not vi tal . Even if only the one side pla te conduc ting the wa ter were cooled , the temperature differ-ential be tween this pla te and the far edge (uncooled ed;;e) of the fuel I

I 14-2

I I plates would not exceed 320*C within 5 minutes. The tempera ture differentials were calculated on the basis of a uniform heat source in the fuel matrix and credit was taken for heat transfer through only two aluminum cladding, neglecting the additional heat transfer I made available by the matrix itself.

Thus, the top feed emergency cooling system is adequately sized and designed to prevent fuel plate melting.

I 14(c) MODIFICATIONS REQUIRED TO THE SYSTEM TO ASSIST THE OPERATOR IN HIS DETERMINATION OF WHETHER TO USE THE TOP OR BOTTOM FEED METHOD A method is proposed for a quick and clearly defined determination of the proper operation of the plenum emergency cooling system in case I of a loss-of-water accident. This is simply based on the fact tha t ,

if operating properly, the plenum system maintains the water level in the fuel elements near the top of the elements. On the o ther hand ,

if the leak should occur in the plenum or other immediate sub-pile I. piping, the vater would be drained from the fuel elements. Thus, the pressure in the plenum is very different for the two cases: the pressure being about 10' to 15' of water greater if the leak is not in the plena.

The relative plenum pressure can be determined readily by simply placing a pressure gauge at the down stream side of one of the I emergency cooling valves. When the emergency cooling line is full of water and the valves closed, the gauge should accurately measure the rela tive plenum pressure.

This concept was checked on October 18, 1966, by placing a compound gauge (-30" Hg to O to 15 psig) immediately down stream of valve DWV-34 I and DWV-35 (see revised Figure 7.1) . The results of the tests are given below:

Emergency Valves I DWV-34 & DWV-35 Settings Compound Gauge Reading (1) Water level at midplane of Open + 10 psig core (LRC-1 reading 50") Closed 1/4" Hg Emergency cooling valves opened and then closed.

(2) Plena drained by opening drain Closed - 26" Hg valves in each plenum. Empty I plena reading: The empty plena reading held steady for 15 minutes before fill was started.

I I 14-3

I Emergency Valves I DWV-34 & DWV-35 Settings Compound Cauge Reading l

(3) Closed drain valves and filled Closed 3/4" Hg to fuel midpoint (LRC-1 reading 50)

(4) Filled to emergency dump level Closed 3/4" Hg (5) Filled to fuel element Closed 1/2"'Hg E transfer overflow level.

E (6) Filled to normal opera ting level. Closed - 8" Hg .

The sequence of readings include Open + 10.8 psig I 2 minute intervals between Closed - 8" Hg readings.

(7) Opened fuel transfer overflow Closed 3/4" Hg and drained to tha t level .

(8) Dumped to dump level. Closed 1/2" Hg (9) Drained plena. Closed 1/4" Hg (10) No change in conditions, one Closed 1/4" Hg hour la ter.

(11) Two hours later opened drain Open _ 10.8 psig valves.

(12) Closed valves. Closed - 26" Hg

.These measurements show that the gauge responds sensitively and uni-I formly to the water pressure in the plena . Items (4) and (8) agree to 1/4" Hg, and items (5) and (7) also agree to 1/4" Hg. The inconsistency for the completely drained plena is ~ 2" Hg, however the reading appears quite reproducible when preceeded by an opening

'I and closing of the emergency valves since item (2) and (12) both gave readings of - 26" Hg.

In these tests the gauge reading for water to the top of the fuel elements is closely simula ted by the fuel element transfer level. This gave a gauge reading of approxima tely 1/2" Hg. The empty plena I which would be the only other condition encountered during a loss-of-water accident would give a reading of - 26" Hg. These two conditions are, thus, clearly distinguishable.

I .

I 14-4 -

These measurements must be repeated for final determination of I expected readings and associa ted limits under normal reactor operating conditions with D 20 and helium blanket.

I Using the numbers generated above as an example, the following procedure would be initiated in case of an apparent loss-of-water accident.

(1) Immediately open DWV-34 and DWV-35.

(2) Af ter 30 seconds, close these valves. and allow compound gauge I- to stabilize.

c (3) If gauge reads between - 13" and - 16" of water, open valves I DWV-34 and 35 and continue use of plena emergency cooling system feed.

(4) If gauge reading is outside of these limits, leave valves 0HV-34 and 35 closed and initiate the replenishing of the I' inner emergency reserve tank by opening DWV-32 and 33.

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I I QUESTION 15.

I Provide infomation regarding testing that was conducted to verify the operability of the reactor vessel fuel handling equipment.

RESPONSE

During the period from April until July 1966, operational tests were perfomed on the upper fuel transfer plug. These tests utilized I the transfer plug with all its pick-up tools and transfer ams, and the reactor upper and lower grid plates. The grid plates and upper plug were placed at their relative elevations in a temporary tower erected in the North ha tchway 'between the first and second floors of B the reac tor building.

The tests perfomed involved (1) alignment of all pick-up tools I and transfer ams, (2) insertion and removal of the dummy fuel element in all core positions, (3) transfer of the fuel element to other loca tions , and (4) trans fer of a 3-1/2" experimental thimble. These tests were also used to train Reactor Operations personnel and to develop I the step-by-step procedure used to trsnsfer a fuel element from all positions.

Figures 15.1 thru 15.5 are photographs of the transfer plug taken during these tests. All operations were conducted successfully.

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f, 5 FIGURE 13.5 FUEL TRANSFER PLUG - 30TTCM "IE' WITH FUEL ELOtC.T L m :J L s

I 1

I I QUESTION 16.

l In order to establish limits on the number of fuel elements which can be handled at any one time, provide the assumptions and re-I sults of a calculation showing the minimum number of fuel elements which could achieve criticality in any geometry, under any con-I ditions of moderation. Discuss the provisions taken to assure that such geometries could not occur in the event of fire in the l

fuel storage vault.

I RESPONSE I

(

16.1 FLOODING THE VAULT l Elementary modified one-group calculations of the reactivity of NBSR elements separated by nine inches (as pertains to the I vault) readily discloses the fact that criticality is inconceivable.

That is, an infinite array of NBSR elements on nine inch centers l

is not even remotely near criticality. Not only is such an array I subcritical, but an infinite array of NBSR elements back-to-back (as in a pool reactor) is also very subcritical. This arises l from the fact that the elements are split and the upper and lower portions are decoupled in light water. Other moderators I were not considered, as there is no credible way in which they could be introduced into the vault. Finally it should be noted l

l that as an extra precaution additional metal supports have been 5 installed in the vault as shown in Figure 16.1. These supports J

will prevent elements from becoming dislodged in the case of some incident such as a fire in'the vault.

j 16.2 FUE" ELEMENT TRANSPORT 1.

I I

Normal refueling of the NBSR requires three fresh elements.

No plans exist to transport more than three from the vault to the reactor at any one time; moreover, these would be transported in a dry condition and lack any credibility whatsoever in going I

j critical during transport. Nevertheless it is of interest to inquire whether some special wet configuration of perhaps six elements could conceivably go critical.

The discussion of 16.1 of course rules out this possibility for the usual back-to-back arrangement in light water; moreover, application of the same elementary theory also rules this out for any conceivable staggering of the elements.

Note should be made of the fact that six fresh 203 .cm elemants contains 1230 ga U 233 or a mass which is only fifty rer 23>

cer.t creater than the mini mm homeceneou s nass (800 nm) of U l

.;-1

o .

in an optimum diameter (1 foot) sphere

• of light water. It re ,

quires a narrow range of spheres or other chunky shapes to achieve criticality with only 1230 gm in light water. For example, an imaginary cylinder whose height is equal to the active height of an NBSR fuci element and in which all of the fuel, aluminum, and light water of six elements is homogenized is well below criticality.

16.3 HEAVY WATER MODERATION Very specia,1 geometries are also required to achieve criti-cality with only 1230 gm in heavy water. The elementary theory applied to the above cited light water cases is insufficient to determine whether six fresh NBSR elements could achieve criti-

{ cality in the NBSR vessel.

known to be impossible in the designed array of seven inch centers.

It does not appear likely, and it is It is therefore safe to conclude that no credible mishandling of as many as six fresh NBSR elements could lead to an inadvertant criticality.

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I QUESTION 17.

Provide a list of manually operated valves whose incorrect posi-I tioning could have a significant effect cn safety. What action will be taken to ensure that these valves are positioned corecctly?

RESPONSE

I The valves listed below could significantly effect safety if they were incorrectly positioned.

To ensure correct positioning, the valves will be aligned according te the master startup Valve Check List which is per-formed before each scheduled reactor startup. This check list is initialed off and kept in a current notebook for the cycle I involved.

I These valves will be tagged with a WHITE TAG (on which is printed in bold type DANGER DO NOT OPERATE) according to ADMIN-ISTRATIVE RULE 11.4.2 of the NBSR Operations Manual.

This rule states that WHITE TAGS are issued by a Shift Supervisor who assigns a number and records their issuance in a " White Tag Logbook." NO OPERATOR OR OTHER PERSON MAY OPERATE I A DEVICE OVER A WHITE TAG WITHOUT THE PERMISSION OF THE SHIFT SUPERVISOR ON DUIY.

j MANUAL VALVES

1. Inlet valve to Emergency Cooling Pump DP-6. (DWV-83)

2. Inlet valve to Emergency Cooling Pump DP-5. (DWV-88)

3. Isolation valve for Dump Valve DWV-9. (DWV-121)

I 4.

5.

Dump line valve at storage tank. (DWV-131)

Isolation valves for valves DWV-32, 33, 34 and 35 between the Emergency Cooling Tank and the Reserve Tank and Plena.

(DWV-266, 267, 268, 269, 270, 271, 272, and 273)

6. Inlet valve to Secondary Cooling Shutdown Pump. (SCV-130)

7. Outlet valve from Secondary Cooling Shutdown Pump. (SCV-134)

8. Outlet valve from Emergency Cooling Pump in Sump Pit No. 4.

(DWV-285)

9. All manual valves in the Helium Void Shutdown System. (HEV-122, 124, 109, 123, 125, 110, 112, 113, 114, 115, 116, 119, and 120)

I I 17-1

a .

I

10. Valves in Low Pressure Backup Air to Helium Void Shutdown System. (LPA-2, 8, 11, 17, 20, 22, 25, 23, 28 and 29)

I 11. Manual valves in 90# backup air for the Automatic Door Seal System. (PAV-1, 5, 8, 9, 10 and 69)

12. Manual valves in 150# air supply. (CAV-2, 9, 12 and 18)

13. Manual valves in 150# air supply to Confinement Door Seals.

(CAV-19, 20, 23, 24, 25, 26, 28, 29, 30, 31, 32, 33, 34 and 35)

14. Manual valves in 150# air supply to ACV valves. (CAV-90, 93, 94, 95, 106, 107, 108, 109, 110, 111, 112, 113, 114 and I 304)

15. Manual valves in 150# air supply to operators of Helium Void Shutdown System valves HEV-1 and 8. (CAV-96, 97, 98, 99, 101 I 16.

and 104)

Manual valves in 150# air supply to operator of Dump valve DWV-9. (CAV-161, 234, 268, 272, 287, 288, 289, 290 and 292)

Refer to the process drawings in NBSR 9 for the location of these I

valves. Also see the attached, revised flow diagrams for the com-pressed air systems.

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QUESTION 18.

Expand your October 1, 1966, reply to our previous question re-garding mechanisms for pressure differential in the confinement building to include:

(a) The extent of leakage which might be expected in the first two hours. Include considerations of heat sinks if credit

[ for these sinks can be quantitatively proven.

(b) The extent of leakage over the course of the accident, list-ing your assumptions and justifications thereof.

(c) Calculated two hour and long term doses. List all assumptions .

and bases thereof.

RESPONSE

p The Response to this question is an extension and modifica-L tion to the response to Question 4 in NBSR 9A. The present re-sponse differs from that in NBSR 9A in that different conditions are considered as suggested by the Division of Reactor Licensing in their letter of December 7. Furthermore, more detailed calcu-

{ lations have now been made which allow credit for certain heat sinks which were not taken credit for previously. Thus, the numbers presented here can not and should not be compared directly to those in NBSR 9A. The conditions which are attendant on the i

maximum creditable accident (MCA) are listed below:

A major rupture of the primary cooling system has

[ 1).

drained all the D 02 (107,000 lbs) from the system onto the process room floor. .

[ 2) All emergency core cooling has failed resulting in a complete core meltdown to the extent that:

a) 1007. of the xenon and krypton (2.9 x 100 curies averaged over the first hour) is released to the reactor building b) 507. of the iodine (2.2 x 106'curies averaged ever the first hour) is released to the building and one-half of this (257. of core inventory) reaches exhaust filters.

3) Emergency exhaust filters absorb 90% of iodine exhausted through emergency exhaust system.

4) No credit is taken for the removal of iodine by the emergency internal recirculating scrubbing system.

5) The barometer is dropping at a rate of 5~. per day for the first day after the acc ide nt .

c) The Lullding leak rate is 24 cfm per inch: the emergency exhaust system maintains the confinement building air at 1/4" water relative to outside; so the building in leakage is 1.257 of its volume per day.

. . 18-1

I Additional effective Icakage (additional pump out rate) results from internal pressure increases due to heat and vapor I sources within the building. The calculation of these forms the j first part of this response and the final dose calculations form the second part.

18.1 SOURCES OF PRESSURE INCREASES I

18.1.1 ASSUMPTIONS. The assumptions on which the calculations are based are given below along with their justifications.

j l

18.1.1.1 D20 remains on floor during 30 day period. This is more conservative than allowing it to be pumped into the j storage tank waere it can not evaporate.

18.1.1.2 Thermal shield cooling failed. If it did not fail,

it would remove all the core decay heat. Thus, it is more con-I servative to assume it fails although the emergency power available to the pumps makes their failure very unlikely.

l 18.1.1.3 The secondary cooling flow stops. This is con-servative since continued flow would greatly enhance the large I

heat exchanger as a heat sink for heat in the air.

l 18.1.1.4 Solar heating continues for 30 days. This gives l the greatest pressure rise.

l 18.1.1.5 Air in process room is well mixed. This assumption I increases the heat transfer rate and assures saturation of the process room air. It is also a logical expectation due to con-vective currents set up by the. warm D20.

l 18.1.1.6 D70 confined to curbed region of process room.

Since the air is well mixed, this has little effect on the tem-perature or saturation of the air in the process room. Further-I more, all the piping that could rupture is in this region.

l 18.1.1.7 The air temperature in the process room and in the other volumes has approximately the same temperature as the wall I surfaces. The wall surface in the process room is about 10 times the D20 surface area so energy exchange with the walls competes very well with the energy exchange between the D 2 0 and the air.

It should be noted, however, that this doesn't significantly impede a temperature rise in the air. Due to the poor thermal conductivity of concrete, the surface temperature rises readily, l

I but the resultant flow of heat into the concrete is a continual drain of energy which is not negligible.

I 18.1.1.8 The tenperature of the secondary water in the heat exchanger and pip in;; in the process roca approximately follotis the ai r ter pe rature . Due to the cood nixing, the heat transfer f ro.a the puddle to these items by convective heating and vapor condensation is quite efficient (see raore detailed discussion in Section 18.2).

I8-2

_ _. _.____________.___.x__

I 18.1.1.9 The core decay heat does not reach the air during I the first two hours. As discussed in Section 1.2 in response to Question 4 in NBSR 9A, the heat must transverse the biological shield first. In two hours, the total decay heat if absorbed by the thermal shield alone would only raise its temperature 35 C.

This is not enough to drive significant amounts of heat through the biological shield and into the air.

18.1.1.10 During the first day, the decay heat is assumed to enter the air. at a rate equal to the core heating rate at the end of the first day. Although the fission product decay rate is greater earlier in the day, most of the heat goes into heating I the thermal and biological shields the first day. Only after equilibrium has been reached will the air heating rate equal the fission decay heating rate. Several days are required to reach equilibrium, so the above assumption is conservative.

I 18.1.2 PRESSURE INCREASE CAUSED BY D70 PUDDLE. The D20 inlet temperature is 37.8 C (100 F) and its outlet temperature is 44.4 C (112 F). The outlet goes directly to the heat ex-I changer where it is cooled to 37.8 C. About half of the water is in the tank and the other half in the external system. Thus, the average D2 0 temperature is about 41.1 C (106 F) compared to a room temperature of 23.9 C (75 F). The total D 2 0 mass is 107,000 I lbs. This large mass and the high specific heat of water makes the D2 0 heat content much larger than that in any of the mechan-ical structure. So only the D2 0 heat content will be investigated.

I 18.1.2.1 Total capacity of heat sinks. The D 2 0 and other heat sources may transfer heat to several heat sinks. The total capacity of these sinks and D 2 0 source are tabulated below to I serve as a guide to the ultimate heat content of the structure.

HEAT SINK OR SOURCE CAPACITIES Air 1.4 x 107 joules / C i Thermal Shield 3.0 Secondary System H 2O 2.8 D0 2 20.0 Concrete Biological Shield 61.0 Floors and Interior I Walls 400 TOTAL 488.2 x 10 7 joules /0 C I For ease in ce >utation the c apac it ie s are stated in terms of the number af j<iles re;;i M to :b y t 'i _ temperature of eacb heat sink by 1 C. The concrete included in the surtary does not i 18-1

a .

I I include the exterior walls or roof. The large heat sink afforded by the H 2 O pool is not included since, under emergency conditions, very little air circulation takes place between that area and the I rest of the building. The spent fuel elements stored in the pool would raise its temperature at about 1-3/4 0C/ day which, as shown below, is less than that in the other sections of the building.

Although the secondary water does not circulate, the amount contained in the heat exchanger and piping within the process room is significa*tt and it is that which is referred to in the summary. Its temperature is 26.7 C (80 F).

18.1.2.2 Effectiveness of concrete as a heat sink. To I determine the effectiveness of the concrete surfaces as heat sinks, the time dependent thermal diffusion equation v 2 7 =

was solved where:

f I T = temperature in C I t = time in seconds c = specific heat = 0.84 joules /gm - C p = specific gravity = 2.3 k = thermal conductivity = 0.0014 w/cm - C The solution gives the time dependent spatial distribution of I the temperature in the concrete for a fixed initial increase in the surface temperature of the wall. The integral over distance in from the surface of one minus the solution multiplied by c is then the total heat which has flowed into the wall in time t.

18.1.2.3 Evaporative cooling. In order to saturate the air, I large quantities of water must vaporize. This results in a significant cooling of the D 20.

18.1.2.4 Calculation of results. As energy from the D2 0 I is put into the air in the form of sensible heat and the latent heat of moisture, the energy is transferred to the walls and the secondary H2 O in the process room. ,

The recirculation fan exhausts air from the process room to the other areas at a rate of 1000 cfm causing additional evaporation and cooling.

The results of the calculations can be summarized as follows.

A f ter two hours all the air in the basement and all the air that I has passed O

of M C.

through the basenent has been saturated at a temperature This is l otte r than the initial D2 0 tenperature of bA ause of the u ce ra l t a al in_, processes 4'..l C

C di scussed above. The e'  :

balance is sur tri:ed beio... .

I 18-4

I l 3 l The volume of the process room minus subpile room is 60,000 ft and the additional air passing through is 120 000 ft3 So the total l

y volume of air saturated at 36 C is 180,000 ft$. The D 20 surface l area is 1200 ft2 The wall floor and ceiling area is 13,500 ft2 and the exterior surface area of the secondary system is 620 ft2, I The change in vapor pressure is that for a change from 50% relative humidity at 23.9 C to 100% at 36 C.

The heat balance requires that the cooling of the D 20 from 410 C to 36 C be compensated by the heat into the walls, secondary l H 2 0, and latent heat of the moisture in the air. Thus, at the end of two hours the heat balance gives: l Energy loss by D 2 0 110 x 107 joules l Energy gains by sinks B Water vapor in air 40 x 107 joules Secondary H3 0 28 i Heat flow in walls to maintain surface at 36 C TOTAL 42 110 x 107 joules So 180,000 ft3 of air has been saturated at 36 C at the end of two hours. This results in a pressure increase due to tem-perature rise of 4.0% and 4.3% for the increased vapor pressure.

I* When the pressure increase is equalized throughout the the net pressure rise is 2-1/2% for the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

building I A similar calculation assuming complete saturation of all the air in the building at the end of one day shows the air tem-perature to be 30 C (due to D 02 only--other contributions are treated separately and added later) resulting in a net increase I in pressure for the first day of 2.8%.

Af ter this the D2 0 is in equilibrium with its surroundings I and only contributes as other sources of heat raise its temperature and hence the vapor pressure. A D 2 0 vapor pressure change of 0.17%

is included as part of the contribution of other sources for each centigrade degree of temperature rise caused by them.

18.1.3 PRESSURE INCREASE CAUSED BY SOLAR HEATING. The dis-g cussion of Section 1.3 of the response to Question 4 in NBSR 9A g is applicabic here and results in a maximum rate of temperature rise of 1.7 C per day or a pressure rise of .68% per day.

18.1.4 DECAY HEAT. The decay heat does not have any effect in the first two hours but is assumed to increase the air (and D20) temperature at the same rate as it increases the biological shield temperature for the remaining t ir e . The heating rate at the end of the first day is 3.8 x 10" w re su l t ing in a pressure rise of 1.F the first day . This is reduced to about .T per day at the end of M days.

l 18-5 .

• e 18.1.5 SUFDIARY. The additional pressure which would have to be pumped out due to internal heat and vapor sources is summarized below:

SUtetARY OF PRESSURE INCREASES First 2 Hours Pressure Rise D02 2.457.

Solar Heating 0.05 Core Decay Heat 0 I

l

+

First Day TOTAL 2.57.

D02 2.87.

Solar' Heating 0.7 Core Decay Heat 1.5 TOTAL 5.07.

Remaining Days 1

D0 2

0 l

Solar Heating 0 77.

Core Decay Heat l Initial rate 1.5 I .7 Final rate Initial 2.27. per day TOTAL Final 1.47. per day 18.2 BUILDING INTEGRITY AND SYSTEM SIZING I 18.2.1 CAPABILITY OF SYSTEM TO HANDLE INITIAL PRESSURE RISE.

The highest pump-out rate is required during the first two hours.

Under the condition of the fastest falling barometer, 57. of the building volume would have to be pumped out in one day plus 1.25%

per day for in-leakage. This would be 0.57. in the first two hours. Combining this with the 2.57. in the summary above would require that 37. of the building volume be exhausted through the emergency exhaust system in the first two hours. This requires an exhaust rate of 150 cfm. Since the measured capacity of the emergency exhaust system is 300 cfm, ample margin is provided, l

i 18-c

l 0 .

I 18.2.2 INTERNAL PRESSURE DROP. Although no mechanism exigts I

l for a really rapid (i.e., within a few minutes) drop of pressure within the building, the internal pressure may be reduced relative to the external pressure in several ways. i 18.2.2.1 Rising Barometer. A rising barometer could in-crease the negative pressure differential across the walls. The l

building is equipped with a vacuum breaker which is capabic of a I

j flow equivalent to 6.37. of the building volume per hour at the negative internal. design pressure of 2.5" H 20. This is clearly I adequate to handle any barometric rise rate since the maximum I

recorded pressure range in the Washington vicinity is less than

57. in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.(1) l 18.2.2.2 Rapid External Cooling. The effect on the in-I j

ternal air temperature of a rapid ccoling of the roof would be delayed several hours as illustrated by the seven hours displacement of the measured solar heating cycle discussed in NBSR 9A in re-I sponse to Question 4. During this time any initial temperature transient would be greatly diffused resulting in a very gradual change in the internal air temperature.

18.2.2.3 Mechanical Pressure Reduction. The mechanical l mechanism which would result in the most rapid pressure drop is very unlikely, but could take place as follows. The secondary system cooling fails at the time of the D 20 flood allowing the air to j warm and become saturated as described earlier. Once the air is warm and saturated, the secondary cooling system is turned on. This I would rapidly cool the heat exchangers and associated secondary piping in the process room to the temperature of the secondary water. Under these conditions there remains the heat and vapor source of the D2 0 which trys to maintain the air warm and saturated I

l in competition with the cold secondary surfaces which try to cool the air and condense the moisture. Since the D2 0 water surface is at least twice that of the secondary system exposed to the air, I the approximation is made that the air remains saturated at the temperature of the D 2 0. Thus the rate of pressure change, de-pending on the temperature of the air and the vapor pressure in it, will depend upon the rate of change of the D 20 temperature.

l The following exchange mechanisms are considered for the secondary system cooling of the D2 0 puddle:

1) Convection cooling of the D20 directly by the twenty-five feet of 20" diameter secondary cooling pipe which would be immersed in the D2 0.

2) The condensation of moisture on the secondary surfaces which would result in evaporative cooling of the D 20.

3) Direct wcond ary coolin; of the air which uauld be re-

• irr J 5- tO D,0.

(1)

Preliminary llazards Summary Report, NBSR 7, Jan. 1, 1961.

I la-7

I In order to calculate the heat transfer, the temperature of the secondary cooling water must be known and an average air velocity I in the process room must be estimated. The secondary cooling water is taken to be at 15.6 C (600 F) since this is the temperature at I which steam heating is annlied to the cooling tower basin. The average air velocity in ..e process room is difficult to estimate, but an upper limit would be the velocity of the air at the in-put registers for the 1000 cfm recirculation system. This velocity I is 45 ft/ min and will be used in the calculations. The heat transfer for item one above was calculated using the Figure 7-11 of McAdams' " Heat Transmission" (2) for convective cooling in water.

I This ga've a total heat removal rate of 5 x 106 joules / minute.

The relationship given on page 249 of the same text for the cool-ing by air flowing parallel to a plane surface was 6used to calculate item three and gave a heat removal rate of .5 x 10 joules / minute.

I Item two turned out to be the most significant factor and was cal-culated as follows. The volume of air impinging on the cold secondary surface per minute was taken to be equal to the air I flow of 45 ft/ min times the projected area of the secondary system.

It was assumed that the moisture content of all air hitting the cold surfaces was reduced to that which would be in equilibrium I at the cold temperature. The projected area of the secondary system is about 140 ft2 So item two contributes a heat re-6 moval rate of 13.5 x 10 joules / min. Therefore the total rate is I 1.9xg07 joules / min. Since the heat content of the D2 0 is 2 x 10 joules / C, this yields a D 2 0 cooling rate of .095 C/ min or about 10 minutes to cool the D2 01 C. This would lead to a pressure drop of about .033%/ minute if the process room were I scaled. When equalized throughout the building this would be a rate of pressure change of about .0033%/ minute (~ .20%/hr).

This rate is about equal to the measured building leak rate at 2.5" I H2 O pressure differential and might not even require the use of the vacuum breaker.

18.2.2.4 I Conclusion. The system and components are adequately designed to handle any decrease in internal pressure.

18.3 DOSE CALCULATIONS I In response to the question concerning recalculation of two hour and long term doses in view of the re-examination of pressure I buildup mechanisms in the confinement building following the MCA, reference should be made to pages 4-6 of NBSR 9A and references therein. The assumptions of the MCA have been restated above and I

the doses are calculated in two ways. First the meteorological conditions are taken to be those accompaning a rapidly falling barometer with good rix4.ng conditions. The exposed persons are assumed to be at the nearest site boundary, a distance of 400 meters. For the len: c m t'o se calculation the persistent <

conditions ace assu- d to prevail in chich the scind b!c , '

a 22. tar 23 ir af th. t; e with an averag sy I

> .t .

o.3 mph. Seutral diftusion coe f f icients are chosen for both c*.

the above calculations.

I (2)

" Heat Tran' -

McGraw-!iilL Wet i c 7, 70,,

Third Edition, Williar H."

7nc, (954,

.g w

I I

I It should be kept in mind that 'the calculations are con-servative since the building leakage is treated very pessimis-P tica!1y. Both secondary system and thermal shield cooling are i denied even though the accident can not be caused by failure of I

j either and simultaneous failure with the MCA is highly unlikely.

(That is, these systems are normally operating and do not have to be brought into operation from standby at the moment of an accident. They are not then in the category of engineered I safeguards. Furthermore, each system has pumps which are supplied from emergency power.) l If the Icakage due to the 2.57. pressure rise in the first p

two hours is added to the leakage due to the falling barometer l and the building in-leakage, the dose to the thyroid for a I

l person, exposed during this initial two hour period would be approximately 48 rad. If the long term dose is recalculated adding'an average 1.87,per day contribution due to solar heating g and core decay heat to the mean in-leakage of 1.257., the resultant g dose would be 73 rad. This latter calculation assumes the build-

.ing recirculation and charcoal cleanup system is ineffective. The combined effective dose would be approximately 120 rad.

I l

These calculations have assumed no credit for charcoal filter reduction of iodine release rates. If a 907. filter efficiency is assumed these numbers are reduced by a factor of 10.

I If the calculation for inversion conditions is redone with the above stated pressure effects the enhanced two hour pump-out rate alters the dose at the point of maximum concentration i at 1050 meters. The dose would be 380 rad. if no consideration I is given to charcoal filters. The total dose for the point at I

l 1050 meters including long term persistent wind conditions and the above enhanced leakage rate would be approximately 400 rad.

The same factor of ten indicated above would apply to reduce this to 40 rad. if 907. filter efficiency in the emergency exhaust system is credited.

In any case these numbers are meant to indicate a con-servctive upper limit to an estimated dose. They show that a considerable margin exists for the operation of the NBSR with reasonably small risk for the safety of off-site public. It I rhould be kept in mind that it is probably not possible for the core to melt with the elaborate emergency cooling, almost certainly not if the primary system rupture does not involve the inlet reactor plena.

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I I QUESTION 19.

4 I We note that the NBSR is in an area of lou seismic probability; however, it is still necessary to analyze quantitatively the po-tential consequences of an earthquake of maximum potential mag-nitude for the area. Include both the confinement bailding and I RESPONSE the reactor plant in your considerations.

The NBSR confinement building was subjected to a rigorous examination of the stresses due to seismic loading. The analysis l I was performed by the engineering firm of Burns and Rte, Inc. of New York. The loads were determined in accordancy with the lat-eral force requirements of the Uniform Building Code of the International Conference of Building Of ficials. For this partic-I ular structure, before taking into consideration the geographic location, the code would require design for lateral forces equal to mass times a coefficient of 8.8 per chnt. The loading tech-I niques of the code are designed tc reflect empirically, the greater scismic acceleration to be expected considering the short period of vibration of the box like structure'and,also accounts for the absence of a moment resisting frame.

The actual configuration of the building lends itself well to resistanc,e to lateral force; the windowless exterior walls of I, 16 to 24 inches of reinforced concrete provide shear walls to accept the loads from the floors as diaphragms of 15 to 18 inches minimum depth, and its near cubical shape gives it equal resist-ance to force f rom any direction.

Under these loading conditions the stresses in all instances I

were found to be approximately one-half of the code allowable stresses for earthquake loading.

Based on historical evidence, the maximum predicted (as opposed to recorded) earthquake i he NBSI area would result in I a lateral g force of .05 to .07 g (roughly corresponding to a 5 to 7 per cent coef ficient). An extension of this philosophy I through geological studies shows a maximum potential earthquake fo rce (as appgd to recorded or historically predicted) in the order of .1 g (coef ficient of 10 per cent). The resultant approximately 14 per cent increase in loadings as compared to I the 100 per cent increase necessary to reach the code allowable stresses still indicates a very large margin of safcty.

I (1) L. Murphy , Chie f, Seismology B ranch, U.

Geodetic Surve S. Coast and (private :camunication Dec. 15, 1966).

I 19-1 I

I I It is recognized that the earthquake provisions of the Uni-form Building Code were established to limit the extent and type of damage which would endanger the occupants, and does not attempt to achieve structures which would survive the strongest earthquake I without any damage. However, considering the approach taken of analyzing for twice the load assigned to the Gaithersburg area as a minimum, and considering the finding that under this con-I servative load assumption stresses are about one-half the Code allowable stresses for earthquake, we would expect the structure not only to be free of the type of damage contemplated by the I earthquake provisions of the Code but also free of cracks which would compromise its function as a containment structure.

The NBSR primary system has been examined for the stresses ,

developed by .1 g carthquake loadings. The system was viewed as consisting of three principle elements; vessel, piping system (including valves and pumps) and main heat exchanger. The most I significant loading result from the inertia of the D2 0 mass in the system. This is largely due to the f act that most of the NBSR primary system consists of large (10,12 and 18 inch) diameter aluminum piping, thus presenting a low dead weight mass with I respect to the D 20.

As stated in NBSR 9 the vessel had previously been examined for forces resulting from .1 g accelerations. The resulting maxi-mum stress from all combined loads (operativnal and earthquake does not exceed the allowable working stress of 6000 psi.

The main D 2 0 leaders, pump connections and valve connections were checked both for D2 0 inertia loads and dead weight mass loads resulting from .1 g resulting from accelerations. The I combined torsional loads (overhung valve operators) and tensile loads on the main D2 0 valve bodies reaches a peak valve of approximately 4300 psi under these conditions. This approaches I but is still within the allowable working stress for those castings.

I Preliminary calculations on the primary heat exchanger, its mountings and connections showed the g caused stresses to be too low (maximum = approx. 1500 psi at mounting feet) to warrent further consideration.

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GENERAL REVISIONS

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I Revise Section 7.1.2, Void Shutdown System, as follows:

7.1.2 VOID SHUTDOWN SYSTDI 7.1.2.1 General

Description:

An additional mechanism for providing significant amounts of negative reactivity is provided in the form of an in-core, helium, gas bubble system as shown in revised Figure 7.2. On an appropriate signal this system will blow helium gas into the reactor core to crecte voids in the moderator. Six helium bottles supply the system with a flow rate of 60 SCni at approximately 25 psig.

This amount of helium would be exhsusted in epproximately 18 minutes; however, if sustained flow is required, the operator using a manual system may release air into the sys tem e t 10 psig backed up by ins trument air. A pressure regulator in parallel with the helium void valve maintains the bubbler s tand pipe at approxima tely 5.5 psig pressure in the static, non-flowing condition, thus assuring the system is charged and ready for opera tion. If the void system is placed into opera tion the reactor scrams and. the modera tor dump valve will open. The reac tor will also scram should the pressure in the bubbler standpipe increase from a leaking valve. See Section 9.5.4.2.

7.1.2.2 Components

Description:

The system consists of six standard 220 SCF helium cylinders arranged in two banks of three bottles each. Each bank is equipped with its own 2200 to 25 psig reducer. Immedia tely upstream of the regula ting station is a remotely ancrn ted isola tion valva , IIEV- 1. Ihck-up sir can be injec ted into the evstem

, m o :h 'irV- 8 .

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20-1

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Helium flow is sensed by an installed orifice and station giving indica tion and low flow annuncia tion on the control room panel.

Control room indication of the " bubbler" standpipe pressure is also provided with a high and low pressure alarm.

Scram protection from a leaking valve is provided by 2 pressure detectors installed in the bubbler standpipe supply line. -

A pressure switch on the bottle header alarms in the control room on low pressure. For information on the design consideration of this system refer to Section 4.6.13.

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Revise Section 7.2.1.8, Spent Fuel Storage Facilities, as follows:

Facilities are provided in the fuel storage pool to receive and store irradiated fuel assemblies, load irradiated fuel assemblies in casks for shipment, cut aluminum end castings from irradiated fuel elements, and to store these castings prior'to disposal. Storage racks and miscellaneous tools are provided.

As fuel is discharged from the reactor to the pool, the elements are placed in a storage rack designed to hold full length fuel elements. The storage racks form a central island in the storage pool. The elements are hung by their pickup heads from brackets as shown in Figure 7.7. In the fueled region of the elements, the dividing partitions are of boral plate. The spacing between partitions is 5-1/2".

See Figure 7.21. A latch bar runs across the front of each fuel element thus pro-viding a secondary means of preventing the elements from falling out of its position.

This bar also prevents the accidental approach of a second fuel element to closer than 3" minimum separa tion from one in the rack. The racks are designed to hold 36 elements in three rows of 12 each.

The boral partitions between elements are added to preclude any possibility of criticality. Even without the boral, calculations indica te tha t a single row of an infinite number of 250 gm elements in an infinite H 2O pool would yield a substantially suberitical configura tion (e f fective multiplication of 2 to 5) .

I For shipping, the fueled sections are cut out of the full fuel element to yield two section per fuel element of about 13" length each. Fi'gure 7.8 shows the design of a rack to hold 48 cut sections 24 full elements. They are stored on sloping shelves. They are separa ted by 9", the minimum space between elements in a vertical column is 6". The vertical columns of shelves are defined by boral I

I 21-1

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partitions and each column holds 4 elements. The spacing between partitions is 5-1/2".

Since the minimum distance (6") between fuel elements in the vertical column is more than twice the neutron migration length, any interaction between fuel elements along the vertical column is very small and the boral prevents significant interaction between adjacent columns.

An area at the east end of the canal is reserved for the loading of cut elements into a shipping cask. The shipping cask is lowered into the canal in this area, opened for loading, lif ted from the pool, washed and removed through an overhead ,

hatchway from the building via truck and trailer.

The transportatf 7n of spent fuel elements to the reprocessing plant will be consistent with health and safety requirements of pertinent AEC and ICC regulations.

The shipping cask will be approved and a Bureau of Explosives permit will be issued for each shipment.

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I I Revise Section 4.3, Control Rods, as it pertains to Shim Arm Shaf ts HISTORY During the course of the shim am test program, the tip of No. 3 shim arm was caught and held by a pipe leading to a bubbler I system test setup. The shim arm was being withdrawn at the time.

Both the shim arm and the shim am drive shaf t were removed from the vessel and inspected. ,

INSPECTION The shim arm was taken to the central NBS shops and set up on a surface table for a couplete dimension : heck. No evidence of bending, twisting, bulging or other distortions were found.

Fortunately, the shim am and pipe mode contact on the solid end plug of the arm. A small dent on this plug was the only evidence of this contact. The alignment of the hub to the shim arm was checked with the equipment used to align these pieces originally. No movement was measurable. Further inspection of the shim am welds and joints was carried out in the Engineering Metallurgy Laboratory at NBS. All welds, joints and surfaces were examined by Mr. John Bennet of the Engineering Metallurgy staff I using the ficrescent dye penetrate technique. No weld or joint failure or other damage of any type was found.

At the initial inspection, the shim arm drive shaft obviously suffered torsion.11 yielding in a region of the splined shaft approximately 4 inches in length. This region I was located immediately behind the hub of the shim am at a point of minimum shaf t cross-section. The stress necessary to cause this measured yield set of 20 is approximately 12,000 psi.

This stress is not suf ficienr. to cause damage to any other portion I of the shaft: this was borne out by the fact that the further inspection revealed no other signs of damage.

REPAIR I Repair of the damaged shaft was accomplished by cutting of the dr.maged tip to a point approximately 6 inches behl.nd the yielded section and joining a new tip to the shaft body. The significant design features of the new tip are the 7/8 inch by I 2-1/4 inch rectr ,ular ke which carries the torsional loads involved in shi, am oper ition ind the 7/8 inch Sy 2 inch lon:

pilot > ?. i f t othith set '!s to cente r tha spline on the shift bo+,

I and which also ca rrie s the r.ajority of the vending stresses resulting from the shift body weight and small eccentricities.

I .

. . 22-1 i

I The maximum loading on the shaft occurs during a scram.

I At the point of maximum acceleratiori, the shaft must transmit a torque of approximately 1735 lb-in plus carry the relatively small bending stresses mentioned above. This loading results in a maximum stress of 8840 psi in both the original and the I modified shaft. The point of maximum stress is the same in both designs; the 1.06 minor diameter of the spline. This last statemen; is made in full cognizance of the stress concentration I factors involved in the two piece shaft. The point of highest stress concentration in the replacement tip design is at the base of the rectangular key. Based on material area alone, the maximum stress at this point is approximately 3800 psi. A stress concentration factor of over 2.2 would be necessary then to reach the 8840 psi stress to be found at the 1.06 spline diameter. A factor of 2.0 and 2.17 is appropriate for the .nodified design.

'A survey of the materials suitable for this task was I conducted with the assistance of Dr. M. R. Meyerson, Chief of the Metallurgy Division of NBS. The material chosen was the precipitation hardenable, austenitic stainless steel A-286.

This is a very tough (high strength plus high ductility) material with corrosion resistance slightly better than the 304 ss used in the original shaft. The properties of this raaterial are shown in Table A below.

TABLE A 304 ss A 286 (Heat Treated Condition)

Tensile Strength 85,000 145,000 35,000 (.27.) 100,000 (.27.)

I Yield Strength Impact Strength 165 ft-lb 64 ft-lb (Charpy V-notch)

Elongation 60L 247.

Reduction in Area 707. 37%

Hardness B 82 C 29 The factors of safety for the point of maximum stress based on the maximum shear stress theory for the original and replacement I

tip are:

E' Original F s

=

S

= '

8840 psi

= 1.358 max Replacement F, = 50,0008840* "

Obviously, the nuch higher strength of the A 286 steel edes I th i s c o.npa r i ,on on f actor of ufety in ti-3: k d. As a f i gu re of merit thcn ha re p l am : mt tip design is t'o u n u i.

joinu in the bod; or ..ift. 'h : ixim n s t re s ; he i .,

I at the centers of the ilat sides of the key slot in the sh.tt.

The 304 ss material at this point is stressed to approximately The factor of safety on the same maximum shear stress 8000 psi.

theory basis is therefore:

22-2

T

• O E 12,000 psi F = = 1.5 s 8000 psi Testing: The two piece shaft was installed in the No. 3 shim arm position of the NBSR and put through 50 full travel scrams.

Inspection of the tip and joint and teasurement of angular I relationship of the spline teeth on the new tip and the fixed teeth on the outer end of the shaft showed no change from the installation neasurements and no other evidence of any movement or level deformation.

Conclusion:

The two piece shim arm shaft tip is completely safe and reliable.

APPENDIX Torsional Shear Stresses were calculated using the following expressions:

For round sections S, = 16T3 (O d

=

T(3a + 1.8b) (2)

For rectangular sections S s I of dimensions 2a x 2b 22 8a b W

For circular segmental S, = T 3

- section Cr where S* = Maximum shear stress (psi)

T = Torque (lb-in) d = Diameter of round section (in)

C = 0.35 for a flat sided circular segmental section r = The radius of a circular segmental section (in)

REFERENCES Roark, R. J. , " Formulas for Stress and Strain," McGraw-Itill, New York, 1965.

Shigley, J. E. , " Machine Design," McGraw-Hill, New York, 1956.

Rot 5bart, !!. A. , " Mechanical Design and Systems llandbook," McGraw-Hill, New York, 1964.

Republic Steel Corporation test data, 1962.

I Materials in Design Engineering, Materials Selector, Reinhold Publishing Co., New York, 1966.

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