ML20085L821

Forwards Final Rept of RO on 831004 Re Possible Operation in Excess of Licensed Power Level

ML20085L821
Person / Time
Site:	05000128
Issue date:	10/28/1983
From:	Feltz D TEXAS A&M UNIV., COLLEGE STATION, TX
To:	Madsen G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
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ML20085L824	List: ML20085L821 ML20086P862
References
NUDOCS 8311030099
Download: ML20085L821 (1)

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{{#Wiki_filter:l go-tW TEXAS ENGINEERING EXPERIMENT STATION i THE TEXAS A&M UNIVERSITY SYSTEM COLLEGE STATioM, TEXAS 77843 [ '~~' N 28 October 1983 M I _. J NUCLEAR SCIENCE CENTER 409/845-7551 Mr.- G. L. Madsen Office of Inspection and Enforcement NOV 3 1983 Region lv 611 Ryan Plaza, Suite 1000 j U.S. Nuclear Regulatory Commission Arlington, Texas 76012

SUBJECT:

Possible Operation in Excess of Licensed Power Level Reported 10-6-83

Dear Mr. Madsen:

Enclosed is a final report of a possible reportable occurrence' observed during operation of the NSCR. This report is submitted in compliance with Section 6.5.2 and 1.29 of the Technical Specifications, Facility License No. R-83 for the Nuclear Science Center, Texas A&M University. Also attached are letters of 17 October 1983 and 28 October 1983 re- { garding extension of the report submittal date. Sincerely, Donald E. Feltz Director DEF/ym Enclosures XA'CoFHiis Bien Sent to PDR T3 J 03co?@W) Y I (y y. =., I RESEARCH AND DEVELOPMENT FOR MANKIND m

1...' y b = Final Report of Possible Operation in Excess' of Licensed Power Level 28 October 1983 RSportable Occurrence [OnlTuesday, October' 4,1983 a high power scram was received during the initial reactor ~ startup 'of the day while banking the_ shim safety control rods after reaching 1 Mw power. Although' operator manipulation of control rods was_ later found to be the cause it was at first thought that ac input line fluctuations may have contributed to the scram since power fluctuations:from campus had been a recurring problem (See Appendix I for scram evaluation). lHowever, during the subsequent reactor startup NSC management _ observed .that safety' Channels #1 and #2 were initially reading approximately 109% and later settled to:106% while the linear channel indicated 1 Mw. (At that time it was determined that

the operator had not closely monitored the safety channels during the first startup'and did indeed initiate 'a high power. scram while banking rods). It was pointed out to the oper-

~ 'ator that fuel temperature and rod positions appeared to be slightly higher than normal, . and the; operator-was directed to reduce power to 100% indicated on the safety channels .until further evaluations could be made. Since the linear channel became suspect, the reactor'was shutdown and investigations were begun. A current source check of the instru-ment indicated no problems,' and therefore it was felt that the detector may have been repositioned accidentally or problems ' existed.with the cabling. The operators then began . preparations for a pool calorimetric to be performed the following day. USNRC inspector 1.es Constable was called to discuss the matter, and during his return call on Wednesday, LOctober 5,= 1983 he indicated that this incident'should be reported. Although temperature equilibrium conditions were established, the calorimetric performed ' on October 5,1983.was not done under xenon-free conditions as is usually the case. The results did, however, agree to within the expected accuracy of the method (i 10%) since an error of approximately -6% was determined (375 Kw based on pool heatup as compared to 400 Kw indicated). ' The detectors were adjusted, and a second calorimetric was per-formed under. xenon-free conditions on~ Monday, October 10, 1983. The results of this calorimetric indicated an error ~of approximately + 4.17 % (416.7 Kw based on pool heatup rate as compared to 400 Kw indicated). Results within i 5% require no detector reposition-ing. However, since this was performed under xenon-free conditions, the detectors were adjusted. The reactor has operated at a maximum 95% power since October 10, 1983 .which.will continueL until completion of the reactor instrumentation study. It should be noted that an evaluation was begun by NSC management on Tuesday, October ~ 4,1983 to determine if the reactor had indeed' operated at a power level in excess of - its licensed 1 Mw limit. A review of the operations log from Monday, October 3 (See ' Appendix 1) indicated that the incident was initiated during reactor operation following - movement of the reactor away from an experimental. device known to have a negative reactivity worth.. The subsequent startup resulted in slightly higher fuel temperatures and safety channel. indications. These differences were noted by the reactor operator. and senior operator but not pursued, and the reactor operated for approximately nine hours 'with both safety amplifiers' indicating'106% /During this same time period the linear chan-nel, considered to be the " prime-standard" power indicator, was indicating a power level. of. l. Mw. =- -

f u ~ pm x-The differences in core parameters were not_ questioned since minor shif ts in safety amplifier readings had been. observed on prior occasion 'due to xenon and experimental conditions ~ Tin the core. However, what was not "?tected by the operator was the fact that based . upon rod position changes a positive. worth was indicated for the experiment just removed. Corrective' Actions (It is felt by the NSC Management t. hat th'ere was an unexplainable disagreement between ~ the linear and safety channel indications that resulted in an overpower of approximately 5% This, however, is within the estimated accuracy of the pool calorimetric, and studies ^ are presently in progress to determine actual reactor power as compared to the linear channel indication.'- Initial results of linearity studies completed thus far indicate that at 1 Mw indicated power the. actual power-is less (~ 4-5%) for both a xenon free and xenon loaded core condition. These studies are still in progress and preliminary results are dis-cussed in Appendix Ill.. Based on these results the actual power of the reactor in this incident was approximately.103L ' The fact that operators involved did not recognize and respond to the conditions is of primary concern. NSC ' management met with reactor operations personnel to discuss and evaluate the incident. In addition the NSC Director routed a memorandum (Appendix 11) to Dr. Carl Erdman describing the incident and what our proposed actions would be. In a meeting between Dr. Erdman,' Don Feltz, and Dale Rogers these plans of action regarding core analysis and operations personnel were discussed at length. A separate meeting was held October 13, 1983 with TEES administrators to review the incident, and proposed actions - were presented.to the facility licensee, Dr. W. A. Porter. Also' attached for review'(Ap-pendix.II) is.a memorandum from the Assistant Director to the Manager of Reactor Oper- .. ations regarding the immediate need for corrective action.' In addition the memorandum 'from the Reactor. Manager to all licensed operators has been enclosed (Appendix II). It should be noted in these memorandums that efforts are being made to stress proper watch-standing techniques. - Emphasis is being placed on use of procedures, proper documenta-tion,' and frequent evaluation of ' plant parameters. Although the incident originated during a shif t change when there is more chance for error due to lack of continuity,~.the two individuals directly involved with the reactor startup on Monday, October '3 have been cc,unseled at length. As part of their immediate upgrade, data obtained from the core study experiments has been made available for their review and ' analysis. ; A followup meeting between NSC management and reactor operations per-sonnel is scheduled for Friday,' October-28. Strong emphasis will be made on the severity of the present problems, and new policies will be established in an effort to increase op-erator awareness and efficiency. The SRO is now required to conduct a facility surveillance

tour and thoroughly review-logs at least every four hours. Operators are to determine experiment r'eactivities following insertion or removals based on rod position changes and compare these values with previously recorded valuds.- It is felt that these requirements

'in addition to those previously mentioned will help to alleviate some of the present oper-ational problems and to generate safer and more efficient reactor operation. p b t I ']

L-APPENDIX I Operations Log 3-4 October 1983 Scram Log Sheet #234, 4 October 1983

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.e 83 n /d/4 /83 Time of Scram 6983 NSC / Date of Scram Log Book: 77 Page M8 O Texas AfM University NUCLEAR SCIENCE CENTER Recovery from Unscheduled Scram I. PERSONNEL ON DUTY: SRO 72:$:// M/ 04/ l)/Af hf" M W HP &NY $$>t/)pte72Ag RO II. CONDITION OF REACTOR PRIOR TO SCRAM: Mode h44s/ M Power Level /MM i Rod Heights:

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SUMMARY

OP OBSERVATIONS: (Alarms, Tripped Meters, etc. ) SS M / 8 4/Af / OA/, SS 8/ Pv4 z ov, .StfE7Y Ib9/k/Re7? SCLAfAf OAl Chart Traces: 'l Fuel Log N Linear Temperature IV. CONSIDERED CAUSE OF SCRAM,*' COMMENTS:#rs h fm ove '7o O/I4detmsvr-8C6u.)CeA) lineo s1 Awax ckk =>el Aucl de 2 ?- @ S4he7Y Cbm*/S, AL Tlsa finaae enormTeN n /Ot * /e SMtTY f AMA)efs cwne Above. IlO *b And mik 200 nossonekr.S 'T5 bank Ands A Powq POVbeY_ fc9)ws taA<

• If a high fuel element' temperature scram occurred, Section VII must be completed and approved before reactor operation is resumed.
• ~

' ' ' ~ - ' _ ~ _ --n.

i. 'u.,e., l Y..WHAT WAS DONE PRIOR TO RESTART'Or REACTOR TO INSURE SAFE AND PROPER OPERATION OP c cvb oP $h*e77 Avos /2UN APsnoJ/e A rr?o d 8F Pye - s.;7),7 r i ~ or eh,A tbmA/ered Rwnx osin/ vp to & r sond sk e,e was enLc/ e w % re h e< nul mas,sa-l. L sesverecJ i,t/ c7Y 95 nit / /~<.wr f/ux nmusemsars meee etAs nac/ ne rse/ JooP! Liew <!w/ ans Kw'n,L! A F/uv fxoknT~ec} thcar A,- V fog VI. CONDITION OF REACTOR AFTER STARTUP Mode Ti>A-} y C7/9,2 Power Level look Rod Heights:

1. 172 0 #272 0 #372 0 #4 72 dTr/ob./ Reg /p 9 Experiments in Core h-

/j A / A operator Signatur T SRO Signature [ scf VII. TEMPERATURE SCRAM EVALUATION Instrumented Fuel Element # TC# / Mode: SS Pulse Description of events leading to scram [ / O Maximum Temperature Observed: Recorder F Do O 4 r f w (0.value from Reactor Log ) x (max.1 fem bserved) = r Evaluation: LSSSw'asexceeded:/Yes No Safety Limit was' exceeded: Yes* No / Evaluation performed by

• If safety limit was /x'ceeded refew to Tech oc 6.3 VIII.

Scram review by Management /,a. i i d. - Date jo/f, f3 ,,Ac e h.am t w 9 x nom. <b4% A e c."" - - ok to aC k *- CG< Q,L. m, - n .uff % u% 4Lx u W CM M a-q s w

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I- = APPENDIX II Administrative and Management Response to Reportable Occurrence of 6 October 1983 4 5 = ,y- -r-e -> ~ - - -- - *,, w w. -*- rre-rrrwe-wr er--wew---- ---m- -t- + - ---r

',c' >o TEXAS ENGIN EERING EXPERIMENT STATION THE TEXAS A&M UNIVERSITY SYSTEM p ~ COLLEGE STATION. TEXAS 77843 7 October 1983 g ) [ "ug g g cEn n MEMORANDUM TO: Dr. Carl Erdman, Head, Department of Nuclear Engineering and Head, Nuclear Engineering Research Dt.!e Rogers, Assistant Director Nuclear Science Center Barry Willits, Manager of Reactor Operations Nuclear Science Center > - il . f'.g FROM: Donald E. Feltz, Director Nuclear Science Center ~

SUBJECT:

Proposed immediate Administrative Respunse to the Reportable Occurrence Dated 6 October 1933 It has been determined by review of the NSCR operations log of October 3,1983 that the individuals involved in the startup of the reactor, following a shutdown to remove an experiment, failed to observe changes in certain reactor operating para-meters resulting in a probable overpower operation of approximately 6% for a period of approximately 9 hours. The probable overpower was detected the following morning and proper procedures were followed in the evaluation and reporting of the incident to the NRC. Although it is impossible to prevent human error, there are measures that can be taken to help prevent a reoccurrence of this nature. The NSC staff must continue to place a strong emphasis on operators observing and questioning changes in reactor parameters and to relate the changes to an operational event or reactor condition. Thus in response to this probable overpower incident the following recommendations are made concerning corrective actions to be taken to prevent a reoccurrence. These corrective measures involve operator training and performance of duties. 1. Immediately place the two individuals involved in the reactor startup after re-moval of the experiment into an accelerated training program emj'hasizing NSCR operation procedures, reactor instrumentation, surveillance of reactor operations, and the assessment of reactivity changes and associated reactor conditions. RESEARCH AND DEVELOPMENT FOR MANKIND

l Memorandum Page 2 2. Provide a similar program for all other reactor operations personnel. 3. Implement immediately a procedure to document the review and assessment of changes in reactor operating parameters resulting from reactvity changes during operation of the NSCR. To evaluate the probable overpower operation of the NSCR and to answer questions concerning differences in actual reactor power versus indicated power for various operating conditions, the following is proposed. Based on the calorimetric measurement scheduled for Monday, October 10, maintain operation of the NSCR at a power level not the exceed 95% of licensed power until the following is determined. Evaluate reactor power vs. indicated power for the following operating a. conditions: (1) Xenon-free core with MRID removed. (2) Xenon-free core with MRID in use. (3) Xenon loadeo core with MRID removed. (4) Xenon loaded core with MRID in use. b. Evaluate linear response of the linear power measuring channel. Establish an independent reactor power monitoring program using the foi! c. activation method. NOTE: If this study cannot be completed within the 14 day reporting requirement, an extension of reporting time concerning the evaluation of reactor over-pvwer should be requested. An incomplete report should be submitted, however, concerning actions taken to improve reactor operator per-f or mance. Following submittal of the 14 day report to the NRC an evaluation of the relocation of the linear detector to the D-9 grid position is proposed. This evaluation should inchde the ef fect, if any, the loading of samples and use of the D-7 notch would have on the linear power measuring channel. This study should also include a linear response evaluation.

Memorandum Page 3 A concerned effort can be demonstrated to the NRC if we take immediate admin-istratise action to prevent a reoccurrence of this incident. The NRC should be made aware of this as soon as possible. The first part of the 14 day report should discuss operator performance concerning this incident and it should also detail the measures taken recently to improve operator performance and surveillance of reactor operation and f acility activities. The second part of the 14 day report should deal with the analysis of reactor overpower and the performance of the linear power measuring channel. cc: Dr. W. Arthur Porter, Director, TEES Mr. Les Constable, NRC L -~

A ~ 3, 10 October-1983 -MEMORANDUM TO: ' Barry Willits, Manager, Reactor-Operations FROM: - Dale Rogers, Assistant Director ~

SUBJECT:

Upgrade of Reactor Operations Personnel - Based on' our discussion with Don Feltz and Dr. Erdman on Friday you should begin immediate plans for operator training. As a minimum I feel the following items should be ' included: 1. Operating philosophy to include. proper use of procedures, log keeping techniques, a questioning attitude,_ more formality, etc. -2. Detailed familiarity _with. Tech Specs. (A written exam for the operations group is probably in ' order). ~

3..lmplement a policy (by NSCO Memo and later SOP) to determine reactivity changes and compare with other core parameters ' following sample handling in the reactor Core.

4 4.~ A review ' of our commitment to - the NRC regarding core analysis following the -- calorimetric' on.10 October 1983 -(e.g. wire and foil meas'urements of flux). Please keep me updated on your progress, and 'if you have questions let me know. / g.

TEXAS ENGINEEltlNG EXPEltlMENT STATION office of the Director October 11, 1983 MEMORANDlM 'IO: Mr.-Donald E. Fe1tz Dr. Carl Erdman ] ' g [/ 3 TIIROUGl: StBJECT: Octcber 3,1983 Overpower of the Reactor 'lhank you for your letter dated 10 October 1983 regarding the action taken following the October 3,1983 reactor overpower. I cannot .over m phasize the importance of quality and continuous training of our reactor personnel. I would like to review with you: (i) the procedures we currently use in training our reactor

operator, (ii) the current-level of training of all of our reactor operators, and (iii) any recanmendations which you have for changes in our procedures and how they cmpare to other reactor operations.

Please schedule this review at your earliest convenience but no later than October 18,' 1983. Mr. Harry Whitmore should also attend this review. W. Arthur Ibrter Director cc: Mr. Ilarry E. Whit 2nore The Texas A&M University System Engineering Research Center College Station, Texas 778434577 e (409) 845-1321

12 October 1983 t MEMORANDUM 'TO:t i D. E. Feltz, Director, Nuclear Science Center FROM: Dale Rogers, Assistant Director, Nuclear Science Center L Ikeview Notes of NSC Reactor Operator Training Programs

SUBJECT:

The following notes have been prepared for the 13 October 1983 meeting called by Dr. Porter to review the latest reportable occurrence at the NSC. 1

1. ' Review of NSC Training Program A. Prior to 1979 there was no formal training program which included scheduled lectures and performance items.. There was, however, an outline on file with the NRC indicating ' reas that ' individuals were to be familiar with prior to the a

exam date.- ' B. 'Following the Three' Mile Island incident in March 1979 more emphasis given to . training'throughout the' industry. CJ Also at about the same time the' NSC.had two individuals fail their SRO exams -(one B.S. and one:M.S. in Nuclear Engineering).- D.' At -that time it became NSC policy to not let anyone take both exams at the ~ same time (RO and SRO). In addition I was to develop a more thorough training . program because of my background as a Navy prototype instructor in the nuclear program. E..; Implemented this program approximately mid 1979,i and since that time we've

had I failure., This new program: included scheduled lecture series by licensed and experienced personnel, performance requirements, periodic progress exams,

-and facility, walk-throughs with NSC ma'nagement. F.' Once. licensed, operators maintain proficiency by attending requalification lectures

and being' examined every 4 months. In addition there are console manipulation it -

, : requirements, quarterly and annually. I do feel, however, that this is an area of possible upgrade in that more extensive training could be scheduled during each .4 month segment. / ' 11.{ Efforts Made to increase Operational Efficiency .'A. - Following appointment as Manager, Reactor Operations in December 1981 I met i with operations personnel to discuss my philosophy on proper watchstanding techniques. +

. p w h f B.. Following overpower. incident and enforcement conference of March 1983 we again met to discuss problem areas. ' C. Efforts being made to keep the operation group aware of overall facility operation this being done by holding periodic meetings and establishing a " required reading" book in the control room. Also see memos for transfer of information: 'I. 18 November 1982 - Log Keeping ' Deficiencies

2. Required Reading Material - 7 February.1983

3. Use of procedures and controlled access to control room - 16 February 1983

~

4. New Tech Spec Requirements - 18 May '1983' D.

Commitment to NRC for Thorough SOP upgrade.. E. Implemented " Trend Analysis I og" on trial basis and now incorporated into oper-ations log. Provides for readings on facility air monitors, reactivity comparisons af ter startup, facility tour and log review by SRO every 4 hours. .1 N 6 d e=%. s ( h-, o/ \\ y V' _ x s l'_ _ M.

4 21 October 1983 HEHORANDUM TO: All Licensed Operators FROH: B. L. Willits, Manager of Reactor Operations

SUBJECT:

Log Keeping / Sample Reactivity / SOP Compliance In an effort to increase operator awareness and develop a philosophy concerning the determination and continual comparison of sample reactivity worths, operators will need to revise or add to their present log keeping responsibilities. These responsi-bilities will not necessarily be incorporated into formal requirements as per an SOP, but will remain a normal routine practice until further notice. A log entry is to be made upon return to full power following the addition or removal of a sample while the reactor is at power. This requirement includes a core movement cway from or against MRID. This entry should state the rod heights before reducing power and the steady state rod heights established after returning to power with the sample configuration changed. It would be most convenient (if conditions permit) to use the regulating rod for this comparison and subsequent sample reactivity measure- ~ent, by adjusting the shim safety rods to the previous average height. The regulating rod height reading can be recorded after switching out of servo to lower the power, rnd taken again after allowing core conditions to stabilize following the return to power. These readings should be recorded in the comments section directly below or beside the log entry denoting the return to power. The operator should then convert this rod haight difference to a reactivity value and record it next to the rod height readings. The reactivity values obtained will be compared with the value listed cn the sample Reactivity Sheet (1 Mw or "at power reactivity") and also with previous ceasurements under similar core conditions. Any significant discrepancies should receive the immediate attention of management. Additionally, operators should concern themselves with ensuring that the Reactor Operations Log is filled out fully. Numerous readings, mainly in the Trend Analysis Cection and sample check areas, have routinely been left blank. Although the readings should be taken hourly while operating or when shutdown while performing certain key evolutions, a brief note can be made and the reading taken sometime in that hour period if necessary. The blank to be initialed following the facili% tour is to include a log review for that four hour period, by the SRO. The SRO will document the performance of these items with his initial. To ensure a more stringent compliance with procedures and to bring about more famil-iarity with these SOP's, we need to be more concerned with referring to a procedure es we perform all evolutions. The appropriate procedure should be reviewed and fol-lowed for both routine and non-routine evolutions.

\\ Memorandum Page 2 I want this requirement to accomplish two major objectives, one of which is, of course, to minimize and hopefully eliminate potentially disastrous S0P violations. The other cajor objective is to note and bring to my attention any procedural inadequacies. We are currently committed to a schedule with the NRC to review and revise a selected number of SOP's (approximately six to seven) each quarter for the next several months. As in the past, I have required the assistance of the HP staff and operations per- '~onnel in all phases of this major undertaking. I feel I have solicited and so far received enough " local review" (by personnel directly affected by these " revised" SOP's) to come up with an interpretable, useful', and updated procedure. Certain areas cf current procedures that are vague or that refer to, or reference other sections of an SOP need to be seriously looked at as areas requiring revision. An S0P cannot possibly detail all conceivable situations. These SOP's need to both allow a certain cmount of room for interpretation and to provide guidelines for the operators to consider while performing evolutions. I want the above objectives to be accomplished by increase usage of all SOP's. e ~

fs s L APPENDIX III Preliminary Results of Core Studies l

~ m v ~ t t y Preliminary Results of Core Studies. Results - r As can be.seen in the summary in. Tables 1 and Il and in Figures.1-10 there appears to - :be a distinct rolloff in actual reactor power relative to the. linear channel indication. J This effect becomes apparent at power levels above 400 Kw. At I Mw as indicated on the" linear channel, three separate alternate. detection methods' indicate that actual reactor n - power ranges from 93.5% to -100% depending upon the core xenon condition and experiment load. F!gure--11 has been provided to show the location of the additional flux monitors: f a fission chamber, pneumatic system sodium' foils,-and a.self powered neutron detector. -Results from_ a gamma chamber located with the fission chamber have not been included because the' calibration and reproducibility of the picoammeter used were in question. (In looking at the response of the fission chamber'the linear detector does indeed seem to be slightly sensitive to vertical flux changes due to rod height differences. In comparing xenon-free vs. xenon ~ $1_.10 (with or without the MRID experimental device) it can be 'seen that.true power (as determined by the fission chamber) is less during the xenon-loaded cases due to-an increased neutron feed to the linear detector causing actual power to be less than indicated. 'It should also be noted that the fission chamber sees a slightly higher neutron flux with the MRID in place due to the shif t in power profile to the. west ~ because of the negative reactivity-insertion along the east face. 1The self-powered neutron detector (SPND) results support the arguments given above in comparing xenon-free and xenon ~$1.10 co'nditions. Because of the detector size and ? location it-is essentially ~ insensitive to vertical flux changes ior the xenon'and xenon free - condition'and sees a lower. reactor power 1under excessive xenon loaded conditions.

The sodium foil results appear to be-sensitive to.-vertical flux changes as evidenced by the fact that the xenon case. values are greater than the xenon free condition ~even though -

actual power is less. This effect is because the pneumatic sample is irradiated at core centerline and the ' vertical shift upward'of the flux causes the sample to see more neutrons for the1 same indicated power level. ' In comparing "with" and."without" MRID results,- ~

the~ xenon free case supports the arguments above in that there is a slight shift' westward This effect is n' t readily obvious under the xenon-loaded condition.

. in the power profile. o h.'-' u

TABLEI Comparison of Alternate Detection Systems Relative to 1 Mw Indicated on the Linear Channel - Xe free Xe ~ $1.10 w/ MRID w/o MRID w/ MRID w/o MRID

Fission.

Chamber 980 Kw 960 Kw 975 Kw 935 Kw - Na Foils - .975 Kw 940 Kw 990 Kw 1000 Kw SPND 970 Kw 960 Kw F I

g. TABLE II Pertinent Core Parameters During'Linearity Tests Xe free Xe ~ $1.10 (Oct. 17, 1983) - (Oct. 20,1933) SS RR Chi Ch2 Temp SS RR Chi Ch2 Temp

No MRID

- 100 Kw 57.0 14.5 202.8 62.5 20.4 214.8 -210 58.0 18.3 24 22 314.0 64.5 18.7 25 23 325.7 400 60.0 21.1 44 43 447.9 67.0 20.3 45 43 461.8 600 62.1 21.5 63 62 559.9 69.5 20.4 65 64 567.7 l 800' 64.5 20.9 84 82 661.2 72.0 20.0 85 82 664.3 1000_ 66.6. 20.5 102 101 735.4 74.5 20.3 106 102 746.8 With MRID. 100 57.0 16.1 205.9 63.5 12.9 217.6 210. 58.0 19.9-24 23 310.6 64.5 17.0 23 23 319.2 400 60.0 22.2 43.5' 42.5 439.9 67.0 18.3 43 43 455.5 600 62.1 23.0 63 62.5 556.2 69.5 18.4 62 64 572.2

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I .i. *- Conclusions The data presented in Table 11 lenos support to the argument that the detectors are sensitive to rod position. The safety channel detectors, however, appear to be more responsive to the vertical shif t than the linear detector since in comparing " xenon-free" vs. " xenon = $1.10" without MRID it can be seen that the safeties read slightly higher. In comparing "MRID" vs. "No MRID" for the xenon-free case it can be seen that there is a feed of neutrons to Safety #2 due to additional graphite at the northeast corner. The xenon case with MRID shows this feed extremely well in that Safety #2 has increased while Safety

1. 1 decreased.

Since the preliminary results imply that actual power is less than indicated, the readings of 105%/106% observed on Monday, October 3 may have indeed been a true power of only approximately 103 %/104 % A comparison of rod heights and temperatures before and af ter the reactor movement on October 3 indicates that the excess reactivity would have resulted in approximately this same power level based on extrapolation of the power vs. reactivity lost curve. The probable overpower of approximately 3% falls well within the experiment error of the calorimetric measurements performed to establish maximum power operation of the NSCR. Exact thermal reactor power measurements are difficult to perform on pool type convection cooled research reactors. This is particularly true of the converted TRIGA reactor. The fuel geometry inhibits the placement of flux wires for determination of the core average thermal neutron flux, thus, the bulk pool calorimetric is the chosen method for calibration of the NSCR power. History of pool calorimetric measurements for the NSCR indicate an experimental error of approximately i 10% This result has been com-pared with other facilities and is in agreement with their results. Operation of the NSCR at nominal licensed power within a variance of 10% does not involve a question of reactor safety nor is there a threat to the health and safety of the general public. The conclusions drawn thus far are based upon the initial studies completed and are pre-liminary only at this point. Further studies are planned and efforts will be made to verify reproducibility of the results. <}}