ML19296B419

Annual Operating Rept 1979

ML19296B419
Person / Time
Site:	05000124
Issue date:	02/01/1980
From:	Curtner A VIRGINIA POLYTECHNIC INSTITUTE & STATE UNIV., BLACKSB
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ML19296B413	List: ML19296B412 ML19296B419
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NUDOCS 8002200535
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{{#Wiki_filter:, T 1 ANNUAL REPORT January 1, 1979 - December 31, 1979 VPI & SU Nuclear Reactor License R-62 Februa g 198Q /- .

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Prepared by: .

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Reactor Supervisor Department of Mechanical Engineering College of Engineering Virginia Polytechnic Institute and State University Blacksburg, Virginia 24061 8002200

TABLE OF CONTENTS I Reactor Operations II Operator Changes III Personnel Changes IV Tabulation of Unscheduled Scrams V Quarterly Scram Time Tests VI Health Physics VII Reactor Refueling VIII Control Rod / Fuel Examinations IX Facility Changes X Reactor License R-62 Expiration and Renewal

I Reactor Ooerations: 2eactor operating parameters for 1979 were as follows: Quarter Jan .!ar Apr-3 me July-Sent Oct-Dec Kilowa t t-ito use 23479 27106 29585 25676 Hours Critical 293.4 301.1 312.8 270.3 Ar Released, mci 37682 43506 47468 41215 Number of startups 50 66 54 54 Unscheduled shutdowns 1 6 18 1 Yearly totals Kilowatt hours 105846 aci Ar 169871 Unscheduled Shutdowns 26 Startups 224 Hours Critical 1177.6 II Operator Changes Mr. Leo Eskin and Ms. M. D. Serrell successfully completed examinations for Reactor Operator Licenses in June, 1979. fr. John Nelson returned to the facility in September, 1979 and successfully completed a requalification program. !r. Vincent Perone left the facility in December, 1979. :tr. ?.obert T. Stone lef t the facility in September,1979 for a 2 year educational leave of absence. III Personnel Changes In September, '.979 Mr. Alan P. Curtner assumed the position of Reactor Supervisor lef t va ant by Mr. Robert T. Stone; ;Ir. Leo Eskin and Mr. Christopher Holland were employed as half time staff reactor operators. IV Tabulation of Unscheduled Scrams Momentary power loss to building 8 Manual 2 Spurious high power 16

The manual scram was caused by a student inadvertantly pushing a remote manual scram button located in the reactor room. The other manual scram was initiated by the reactor operator when an excessive AT and high moderator temperature was observed, refer to attachment A for details. The numerous spurious high power scrans have been caused by a momentary jump in power indication on the power range #2 instrument. All of the circuitry in this channel has been replaced with spare modules, with no change in the observed problem. A review of the scram log for previous years of significant construction activity on campus show a similar number of spuricus scran.s during such periods, indicating a possible A.C. power disturbance as the cause. These scrams occurred during the spring and summer aonths while heavy construction work was under way in the vicinity of Robeson Hall. The spurious scrams abruptly stopped in late fall and winter when the construction work was terminated. V Ouarterly Scram Time Tests (seconds) Jan . tar Aor-Jung July-Sect Oct-Dec Average Ouarter

                                          .39       .41          .57          45       46 Safety ill Safety #2                       .39       .49          .51        .56       .49
                                          .37       .42          .45         .50       44 Shim-Safety VI Health Physics Area surveys and wipe tests were made on a minimum of a quarterly basis.

There were no significant changes in observed radiation levels during the The only radioactive waste released to the environment was Ar through the ventilation stack. The total amounts released per quarter are shown on the operations summary. VII Reactor Refueling During March, 1979 a complete fuel element (12 plates) was replaced. Also inner fuel elements vere swapped with outer fuel elements within the core in an attempt to gain a net positive reactivity increase. The set basic excess reactici y was measured at .317% AK/K. VIII Control Rod / Fuel Examinations The control rods and reactor fuel were inspected during the week of the reactor refueling. There was no evidence of cracking, pitting, or unusual wear or corrosion. Control rod worths, stringer worths, and reactivity input rates were measured. No significant changes were noted.

IX Facility Changes A. Neutron Detectors In February a CIC detector located in the shield tank was connected to the Keithley micro-nicroammeter; and a UCIC detector located in the thermal column was connected to the power range channel #1 instrument. This completed the action of moving all detectors from the core top region to the shield tank detector tubes and thermal column region. (see attachment 3 for background & details) Detectors located in Thermal Column experimental ports: UCIC gower range channel #1 instrument UCIC - power range channel #2 instrument fission cham'aer - source range instrument Detectors located in the Shield Tank: CIC - intermediate range instrument CIC - Keithley micro-microammeter B. Reactor Control Console A proposal to add instrument racks to the reactor control

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console was reviewed and approved by the Radiation Safety Committee. (see attachment C). In 1979 the following instruments were installed in the racks:

1. A CCTV monitor

2. Large digital clock

3. Two Rustrak chart recorders for recording radiation and fission products in the stack

4. Reac time ter

5. A second annunciator alarm panel

6. Fission product monitoring instruments for stack ef fluent Safety Evaluation In the judgement of the Radiation Safety Committee, the change does not: 1. increase the probability of an accident or malfunction of equipment important to safety previously evaluated in the Safety Ana'ysis Report, or; 2. create the possibility of an accident or ro.J2cretion of a different type than any evaluated previously in the Safety Analysis Report, or; 3. reduce the margin of safety as defined in the basis for any Technical Specifications.

C. Reactor Safety System A proposed modification to the reactor safety system was re-viewed and approved by the Reactor Committee. (see attachment D.) The change was installed and tested satisfactorily in November, 1979.

D. Back Flow Preventer Valves Back flow preventer valves were installed on the secondary cooling system inlet by the Physical Plant Department. Subsequent operational problens resulted. ( see a ttachment E.) E. Fission Product !!onitorine System for Stack Ef fluent This monitoring system has been installed. Tests and cali-bration are still in progress (see attachment F.) X Reactor License R-62 exoiration and Renewal The ten year R-62 Reactor License expired on November 16, 1979. A letter of application for license renewal was mailed to the NRC in October, 1979. Supporting documentation required for renewal is being prepared and will be forwarded to the NRC in the future.

• Attachment A VIRGINIA'S LAND GRANT t/SIVER$1TY me

 . h ,,j         VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY M t*'
   ~                                                                                      Slac tsburg. I trgrma :st%l Ontc or Occtr4Ttos4L H e = t.ru a m o d a r m (703) 951 4 *75 MINUTES RADIATION SAFETY COMMITTEE Interim Meeting 3:30 p.m.                                  108 Robeson Hall                           May 7, 1979 PRESENT:       A. K. Furr: R. D. Mogle; S. L. Meyers; A. P. Curtner; A. Robeson; R. T. Stone; J. R. Craig; J. L. Shotts; C. E.

Polan and T. F. Parkinson, Secretary. ABSENT: R. A. Teekell, Chairman; John W. Cure III; G. P. Beck; H. L. .iood A. The meeting was chaired by Dr. A. K. Furr vice Dr. Teekell who is on annual leave. The purpose of the meeting was to review an ab-nomal occurrence in reactor operations , on ".ay 7,1979. B. A. P. Curtner reviewed the abnomal occurrence. App roximately 15 minutes after the reactor was stabilized at a power of 100 kw, he observed that the coolant outlet temperature was abnormally high and was rising faster than normal. A LT value of approximately 500F vas noted. All other instrumentation gave normal readings for 100 kw power. R. T. Stone then checked the mechanical thermometer located in the process pit and confirmed the abnomal value of AT. He also checked the coolant flow rate and confirmed that it was within the normal range. At this point, the reactor was shut down so that the coolant outlet temperature did not reach the high alarm level of 1650F. C. Some discussion ensued on the abnomal occurrence and it was agreed that the most ',srobable cause was a misalignment of tha moderator dump valve which allowed some by-pass of the flow. D. R. T. Stone proposed the following procedure to check the proper operation of the dump valve: A sequence of 10 tests shall be perfor=ed with the reactor shut down. Proper operation of the dump valve shall he confirmed over a temperature range frem approximately 85*F to ll5 0F. Each test shall last at least 10 minutes.

Radiation Safety Co=mittee Interts Meeting Mcy 7, 1979 Page Two E. A. K. Furr pecposed that the standard maintenance procedure should be performed on the power supply for the du=p valve clutch voltage. F. R. D. Mogle proposed that prior to full power operation, the reactor should be operated at 50 kv and that a heat Salance check should Se performed. G. The committee members then di4 cussed whether the incident should be reported to the NRC. It was agreed that the NRC would Se notified by telephone with a follow-up letter. R. T. Stone e=phaa12cd that the applicable emergency procedure was fo,llowed and that no Tec. finical Specifications were violated. H. It was agreed that the reactor staf f should review other possible corrective =easures that =1ght be taken to avoid a tepetition of the incident and that a report should be given at the next regular meeting of the RSC. I. A motion was made and seconded that the above reco=mendations were approved and that, if all checks proved satisf actory, teat normal reactor operations could be resumed. The motion passed unanimously. The meeting adjourned at 4: 30 p.m. Respectfully submitted, l ~ T. F. Parkinson Secretary

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May 9, 1979 Dumo Valve Leak Test ?asults and Reactor Recoverv After Unscheduled Scram on Mav 7, 1979 3ackground On May 7,1979 at 0938 the reactor was manually scrammed by the reactor operator when an abnormal AT (dif ference in inlet and outlet coderator te=perature) of = 500 F was observed. The reactor was operating at a steady state 100% power when the abnormal temperature was observed at approximately 0934 The outlet temperature was at 160 F and rising with inlet at 110 F. Moderator coolant flow was normal at = 21.5 gpm. Indicated power

     = 157 Kv.by heat balance calculation from these parameters was read normally However, for 100%:at the same time the nuclear instrumentation Keichley : 3.2x10 -5 amps Power CH.1 :     99*

Power CH.2 : 96* Process pit radiation monitor: = 100 cr/hr All other area radiation monitors read nor= ally. With a 57 Kw discrepency between the heat Ealance calculation and installed redundant nuclear instrumentation, a malfunctioning remote temperature indicator was suspected.

• The reactor assistant was insediately sent into the reactor cell and read the mechanical dial thermometers:

Dial Thermometer Console Monitor Inlet 111 F Outlet 112 F I 164 F 163 F When the dial thermometers confirmed this was the actual temperature leaving the reactor, the decision was made to scram the reactor. At the time of scram the outlet te=perature was 164 F and still rising . This indicated awas the temperature heatfirst up observed. rate of greater than 1 F/ min. from the time The reactor assistant also checked secondary pressure guages and valve lineup; all tndicated normal. Af ter the found reactor to be scram a complete valve lineup was performed and satisfactory. coolant flow calibration check performed.The du=p valve was closed again and a The check was satisfactory: Indicator on console Actual 21.6 gpm. 21.7 gpm. An interim RSC meeting was called together at 1530 May 7, 1979. Af ter extensive discussion, the conclusion was reached that the

du=p valve by some means had drif ted off its shut seat alleving a substantial amount of coolant water to bypass the core tanks directly to the du=p tank. The RSC outlined a test procedure prior to restart of the reactor. Test Procedure L. Perform a =inimu= of 10 leak test rates of the dc=p valve over a varying te=perature range.

2. Perform maintenance checks of the du=p valve clutch power supply.

3. If steps 1 and 2 are satisfactory, take the reactor critical and perfor= a heat balance calculation at 50 Kw prior to resu=ing full pcver operations.

Test Results

1. Performed leak rate tests on the dump valve, data Eelow:

Time Water he12ht (in.1 Run # Begin End Temp. ( F) Begin End LeaE. rate (gal /Erl 1 0911 0921 83.0 47 1/8 47 1/8 0 2 0927 0937 92.5 47 1/8 47 1/8 0 3 0945 0955 99.0 47 1/8 47 0.7 4 1012 1028 106.0 47 46 3/4 1.0 5 1041 1051 115.0 46 7/8 46 5/8 1.5 6 1101 1111 121.0 46 5/8 46 3/8 1.1 7 1128 1138 125.0 46 5/8 8 46 5/8 G 1147 1157 130.0 46 5/8 46 1/2 0.7 9 1207 1217 132.0 46 5/8 46 3/8 1.0 10 1229 1239 132.5 46 5/8 46 1/2 0.7 Notes:

1. Tests perfor=ed with core tanks filled, reactor coolant pump off and its discharge valve shut, and ion exchanger loop isolated.

2. The dump valve remained closed for runs 1 thru 6. It was opened and closed between runs 6 and 7, then lef t closed for runs 7 thru 10.

3. The water height was read on the high level scram sight glass.

The dump valve was reopened after 10 runs. The reactor coolant system temperature was lowered to 93 F. The dump valve was reclosed and a continuous one hour leak rate test run. Data below: Time Water height (in.) 1305 47 1/8 1315 47 1325 46 15/16 1335 46 7/8 1345 46 3/4 1355 46 11/16 1405 46 11/16 Leak rate: 0.44 gal./hr.

                                      ~3 ~

2. Performed checks on solid state relay cutputs as per tee Sailey Instrument Manual, all checks were satisf actory.

3. A heat balance was performed at 50 Ks:

Heat balance calculation : 50.3 Kw Power range channel 1  : 51 Kw Power range channel 2  : 50 Kw .

                                                   ~#

Keithley Instrument  : 1.6x10 amps (100%-3.3x10-5a=ps) Conclusion

1. All tests were performed satisfactorily.

2. The indicated leak rate of the dump valve is not sufficient to cause the temperature rise which required manual scram.

3. The dump valve opened partially from a mocentary p ower loss or voltage drop from the output of the solid state relays.

4. The decision was made by the reactor staf f to continue nor=al full power operations considering the results of tSe preceding tests.

. Attachment B e

                                               - 'I l'. b r oa r y ! !, 19/'l l'rocedure for Shif ting the seithly;_ instrument CIC Detector from the cere ten reglec to the south instrument tui'e minen t ed in the s!.leid tank

1. "l a e installation ptocedure as outlined in '.diation Safety Committee meeting minutes for Septecher t 'i . '.978 will be followed, e;eept for the reference to an '-li nt leal detec te r in the thermal column, for this proceduit , .we the losation of the identical detector to one located in the shield tank instead of the thermal column.

2. ihe following is data accuuulated from a spare CIC dercerar that has been installed in the shield t.uik since December 1, !"!..

3. T..e data below indicates a maxieum of 5/. irlance in current out of the CIC detector at 100 K'a' over .i two montli period.

The data also ind icates a :raximum o f 5 va r i ince in the curriut .> a t of the kelthly CIC detector ovei the tame tire period. Keithly Shield T. ink Date Tine Instrument CIC Detector 12/3/78 1058 97 amps 39 o amp 3 1158 93 40 1258 92.5 39 1358 9 2. , 39 12/6/78 0931 91 40 1131 93 39 1231 93 39 1431 94 40 12/11/78 0909 92 39 1009 92 39 1109 92 39 4209 92 59 1309 92 39 12/12/78 0900 92 39 1000 91 41 1100 95 41 1200 96 41 1300 96 41 1400 96 41 12/19/78 1002 92 39 1102 92 39 1202 92 39 1/3/79 0959 90 39 1059 92 40 f 1159 91 38 94

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Vetihlv 'ib ' e l al T i, Date Time I t'. a r r omt 's t C I' ;ietect<r 1/4/79 t19C4 'l l e r- 41 , , , , , . 1004 il n

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                                                       'O 1/8/79     1057         91                      .o 1157          33                   45 1257         44                    41 1357        4 '.                   40 1457          1                  40 1557          35                   40 1/15/79   0903         '82                    39 1003          i 's                  40 1103         96                    40 1203         96                    40 1303        96                      '0 1403         16                    40 1/19/79   0939        95                     40 1039        95                     40 1239        15                    41 1339        96                    41 1439        9'i                   40 1/25/79  0914         97                     39 1014        97                    40 1L14        97                    41 1214        97                    40 1314        97                     41 1414        97                     40 1/ 30/ ') 0959        92                     '
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                 !!59        94                     39 1259        9 '.                   r 1359        92                     lo 1459        95                     ~; 9 2/12/79    0902        92                     19 1002        93                     39
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               '. t x Cata be!av indicates i m i x i num o f .'* s .. . int e in t i.r r .. n t out at t h I'JI C de tev: t or .i t 100 K',.* o v e r a < ine ncnth nerfo.!.

A n ' t. 9 .artance of M in current out ..t the V. . i t ', h I I' detector e iurred .)ver this sa.e ti c rer!..l. . o , p nbod-(y 4, y pr ,o u N - - -

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   ,s-Keithly      Thermal Column Date     Time Ins t r ument       ICIC 1/10/79  0921    91o . imps       290 ,, rps 1021    44               288 1121    93               287 1221    96               2Si 1321    96               286 1/11/79  0903    93               235 1003    95               285 1103    95               235 1203    95               283 1303    95               283 1403    95               285 1/15/79  0903    92               285 1003    95               283 1103    96               283 1203    96               285 1303    96               283 1403    96               283 1/16/79  1303    97               283 1403    97               283 1503    97               283 1/17/79  0853    93               284 0953    97               281 1053    97               283 1153    97               283 1253    97               283 1353    97               282 1453    97               284 1553   97               285 1/22/79  0908   92                287 1008   95               289 1108   95               285 1208   95               283 1308   95               283 1408   95               289 1/30/79  0859    92               286 0959    94               235 1059    94               285 1159    94               286 1259    94               286 1359    95               286 1459    95               289 2/6/79   0923    93               287 1023    93               287 1123    93               289 1123    93               286 1323    93               286 1423    93               287 2/8/79   0928    94               288 1028    94               288 1128    95               288 1347    95               288 1447    95               288
            ?/9/79   0851    93               288 0951    95               288

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V ld {j)Qij yWa; l - , y --eg Proposal to Shift Locatien of the Power and Startup Range Detectors from the Top of the Reactor Core to the Ther=al Colu=n Stria.gers Rackereund The VP15SU Nucicar Reactor was originally licensed to operate at a maximum rower o f 10 A"J. In 1969, the license was amended to per=it operation at a maximum pcwer of 100 iG. In effecting this change, the neutron detectors were not re-located, but were shielded with cadmium to reduce their output by a factor of apprcxi=ately ten. Since 1969, the most significant reactor maintenance pecblem has been repairing tne neutron detector cables and fittings that deteriorated due to the high radiation fields in which they are located. The greatest source of radiation exposure to reactor personnel has been received in carrying out this maintenance. Accordingly, studies have been made of the feasibility of relocating the neutron detectors in regions of reduced radiation intensity. In June of 1977, the intermediate range detector was relocated in the bulk shield tank of the reactor, in accordance with procedures approved by the Radiation Safety Cocnittee, and has operated

  ,a tis f ac torily since that ti=e. In ! arch, 1978, a fission chamber and an un-co=censated ion cha=ber were installed in the stringers in the graphite thermal

clumn to experi=entally dete:mine the feasibility of locatins; permanent detectors in tnat location.

  'rocosal it is proposed to replace the source ranga neutron detector now installed
  .a t,e core top area (southeast corner) with an identical fiasiun chamber installed in ene graphite thermal column four inch removable stringer located at the midplane of the reactor core vertically, and to replace the Fever range One ancor..pensated ion chamber with an identical detector located similarly to tne dission chs=ber.     (See attached drawings.) The Source range detec tor change procedure is greatly si=plified by virtue of five months of operation with
 .. n identical indicating circuitry to that currently used as the reactor source cange instrument. The chamber current for the power range instrument has been
  'bserved and logged for five months, facilAcating recalibration of the power rang instrument to accocadate the detector substitution.

D i sc us s ien The results of five conths of study indicate a much greater correlation between the power indications f rom the ther=al column detectors and the calorimetric reactor power than has been observed with the core top detectors. Additionally, the detector stability over long operating time is greater. The signal-to -noise ratio of the fission chamber is significantly greater than with the core top

  !ctector, and the signal level is approximately 40 times that of the core top vtector. Approxi=ately 1.5 decades of overlap is provided between the ter ediate Range indicatien f rom the Shield tank detector and the Thermal column eurce range deccetor.

Several advantat,es bect =e apparent in this relocation. The position of the letectors is such that chay will not be disturbed during annual in-core maintenance, _.a tne absence of esd=ium rhielding will mean less signal level change if detector

  .aintenance is required. T'e moisture-free environment means sta51e operation

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       '/;[ can    be provided without requiring sealed environ = ental char.bers for the detectors and connectors.
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Safety Analysis One passible safety questien was considered - does :he new location allew an experiment to shadow the detector. There is no space between the reactor core graphite and the thernal column graphite other than that occupied by a two inch lead gamma curtain. Therefore, it is not possible fer an experiment to shadow the detector unless the exreriment is installed in the same stringer hole as the detector. This will be forbidcen. Accectance Criteria The acceptance criteria shall be that the detectors installed in the graphite thermal column be identical to the the detectors currently in use, and that the performance of the detectors in the new locations provide signal levels greater than or equal to those currently provided, with equal or better stability. The dats accuculated over the five month observation period indicates that these criteria are net, installation Procedure The installation cf the thermal colu=n detectors as operating instrur.entation shall be effected by:

1) Installation of conduit or other such suitable shielding to protect the signal cables to the thermal colu=n.

2) Evaluation of the accumulated data to assure acceptance criteria have been met by all licensed Senior Reactor Operators at this facility and by one consultin; member of the Radiation Safety Co=mittee.

3) Calib ation of the Power Range One instrument to provide 100*. indication at the current level observed from the thermal colunn UCIC at 100*

calorimetric power.

4) Connection of the thermal column detectors and performance of a reactor flux calibration at a reactor power of 10P.,

5) A mini =um of 10 hours operation at power levels less than 90% with no observable deviation from normal indications on the replaced detector instru=entation.

The proposed change may be ef fected prior to approval by the Reactor Licensing Branch of the U.S. Nuclair Regulatory Commission in accordance with 10 CFR 50.59 if, in the judgment of the Radiation Safety Committee, the change does not: 1) increase the probability of an accident or malfunction of eqaipment important to safety previously evaluated in the Safety Analysis Report, or; 2) create the pos-sibility of an accident or malfunction ef a dif ferent type than any evaluated previously in the Safety .\nalysis Report, er; 3) reduce the margin of safet' as cefined in the basis for any Technical Specification.s.

Attachment C Proposed Addition and Modification to Reactor Control Console Instrumentation 32ck2round Extra instrument rack space at the reactor control censole is needed for existing instru=entation, additions and =cdifications that will be necessary in the future for the proposed 500 Kw upgrade in power.

 )tocosal

1. Two instrument racks (figure 2) will be constructed and added to the reactor control console (figure 1).

2. An additional annunciator alarm panel will be added above the existing panel to provide additional alarm indications.

a. The following existing indicating lights will Se converted into annunciator alarms since they are not

                       " eye-catching" enough for alerting the reactor operator of an abnor=al condition:

1. water in process pit

2. secondary water pressure low

b. Possible additional annunciator alarms:

1. low secondary water flow

2. Fission products in the stack

3. Low level, cooling tower du=p cank

4. High pri=ary water conductivity

5. High secondary water conductivity

3. The existing micro processor based reactimeter will ha relocated from its temporary wooden box on the floor to a space in the instrument rack.

4 A seccadary cooling system control and monitoring panel vill be added for 500 Kw operation. 5. A chart recorder and instrumentation will be added for the proposed fission product monitor that will be installed to sample stack effluent.

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t O PROPa5ED ADD \T iO Ms & C.'-IA NG ES

Attachment D Proposed Modification to the Reactor Safety System Back2round During the past year (1979) all preventive maintenance and calibration procedures have been rewritten into a core detailed format. This rewrite also included a detailed calibration procedure for the Nuclear Instrumenta-tion Drawers. A deficiency in the bistables that provide power and period scram signals to the scram logic instrumentation and reactor safety system nas been discovered during this procedure rewrite exercise. The Bistable Trip A-25 "odule has a potentiometer for adjusting the trip setpoint level for the scram signal (typically set for 120% over-power scram in the power range drawer, and 5 second period scram in the intermediate range drawer). This part of the bistable functions normally. However, what is lacking in the design of the bistable is the ability to adjust the reset point. This inability causes an inadequate deadband. With-out this deadband adjustment capability the following event could occur: A high power or period scram is initiated - control rods commence insert-ing into the reactor - the scram condition immediately clears (1 to 800 milliseconds after scram initiation) - the control rods relatch prior to full insertion due to the clutch voltage reenergizing. The safety rods could stop inserting at any point between their 16 to 0 inches of travel. With the high power bistables set at the 120% power trip point (set during the calibration procedure), the observed reset point varies between 119.5% and 120%. The 5 second period scram bistable resets between 5.2 and 5 second period. This is not an adequate deadband since control rods could turn power or period rapidly enough such that the affected bistable resets and re-latches the control rods prior to full insertion. It must be stressed that this inadequacy does not pose any hazard to public health and safety, nor does it reduce the ability to maintain the integrity of reactor components and systems in the case of an abnormal occurance. However, by all standard reactor engineered safeguards systems design, the instrumentation has the inherent possibility of not performing as it was intended. Therefore this deficiency should be corrected. Proposal A change in the Reactor Safety System circuitry will be made to correct this problem (figures 1 and 2) . A wiring modification will be made to the existing back-up scram relay. If a momentary high power or period scram condition occurs as described earlier, the relay will sence this and " lock-in" the scram condition, preventing a possible relatch of the control rods.

Safety Evaluation In the judgement of the Reactor Committee, the change does not:

1. increase the probability of an accident or malfunction or equipment important to safety previously evaluated in the Safety Analysis Report, or; 2. create the possibility of an accident or malfunction of a dif f erent type than any evaluare' previously in the Safety Analysis Report, or;

3. reduce the mark _.. of safety as defined in the basis for any Technical Specifications.

Installation Procedure , 1. ueenergize the reactor control console by recoving the key, and danger tag off the key switch on the console.

2. Perform voltage checks at appropriate circuit points to be altered to insure power is deenergized.

3. Install the scram reset relay K-32 in a convenient location in the control console.

4. Disconnect the wire between J3-4 and pin 2 of relays K-10, K-11, i;-14

5. Connect hook up wires from pin 2 to pin 6 on K-is, J.-11, and G-14 respectively.

6. Connect a wire between TB9-21 and pini , 6, andil on K-32.

7. Connect wires between K-32 pin 3 and 36-1, -32 pin 7 andJ6-2 K-32 piu 9 and J6-2.

8. Connect from TB9J4 and a spare normally open contact on scram reset switch Pb-7.

9. Connect a wire between the other contact on PB-7 and K-32 pin 10.

10. Connect a wire between K-32 pin 2 and tb9-84

11. The installation is complete, perform the Test Procedure.

Test Procedure

1. Perform the reactor precritical checks.

2. Place power range channel #1 in test and place the test signal function switch to the trip cal. ?osition.

3. /Wjust R1 on the power range test circuit until 1217. power is indicated.

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4. Return the power range drawer to operate and place the test signal function switch to 100% position.

5. Shut the dump valve and withdraw Safety Rod No. 1 to 16 inches.

6. Place power range channel #1 in test.

7. .tapidly toggle the test signal function switch from 100% position to trip cal. and then back to the 100% position.

8. Insure Safety Rod No.1 fully inserted to 0 inches and the dump valve fully opened.

9. ?,ese t the scram condition and repeat steps 2 thru 8 for power range channei #2.

10. Return the power range drawer to operate.

11. Reset the scram, shut the dump valve and withdraw Safety Rod No.1 to 16 inches.

12. Adjust R10 on the Intermediate Range Test Circuit fully counter clockwise. Place the I.T. test signal function switch to the trip cal, position.

13. Adjust R10 on the test circuit until the period meter reads 5.5 seconds.

14. Rapidly adjust R10 from 5.5 second period to 4.9 second period and then back to 5.5 second period.

15. Insure Safety Rod No. 1 fully inserted to 0 inches and the dump valve fully opened.

16. Place the Intermediate Range drawer to operate and reset the scram condition.

17. Test of the safety system modification is completed.

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TM a D NEW BUSINESS T G. Proposed Modification to the Reactor Safety System (see Attachment H).

   %            The change to the system was presented for review and approval. It
   $            was reco= mended to clarify the statement that the inadequacy in the
   !?           existing system does not pose a hazard to the public health and safety k            nor the ability to maintain the integrity of reactor systems or com-
   $1           ponents.
   %                 The rate of any power excursion is directly proporticnal to the H            amount of positive reactivity inserted.

f, If a very rapid power excursion occurred the power and period overshoot would be enough for control rods to adequately insert and p FL turn power prior to the bistable trip A-25 codule reseting. d If a slow power excursion occurred power would still be turned o around even by a partial insertion of control rods caused by the pre-f T mature resetingof the bistabletrip A-25 module. A motion to approve the Safety System Modification and for the reactor staf f to implement it passed unanimously. m

- Attachment E COLLEGE OF ENGINEERING

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[g[$$ ) VIRGINIA POLYTECilNIC INSTI'lTTE AND STATE UNIVERSITY Bladsburg. l'irginia l%b1

• Niecusa Acnymow Av uus Lison arons January 4, 1979 Mr. James R. Cle= ens Physical Plant Department Mechanical Services 6A Maintenance Building Campus

Dear Mr. Clemens:

On July 30, 1979 Physical Plant installed 2-parallel lh inch back flow preventer valves on the 2 inch inlet cooling line that supplies water to the heat exchangers required to remove heat generated in the operation of the nuclear research reactor located in Robeson Hall. The installation of this size back flow preventer has adversely affected the operation of the nuclear research reactor when operated at 100% power. ne back flow preventer valves have sufficiently reduced the pressure and flow of secondary cooling water to the reactor such that temperature limitations imposed on us by our technical specifications were reached requiring us to reduce power to less than 95% on certain days during the remaining summer months. 100% power operation is re-quired for performing Neutron Activation Analysis, a major on-going analytical service used by many researchers on campus and numerous other outside organizations arcund the country. I anticipate we will not be able to operate the nuclear reactor at its full potential to provide needed research services during the months of May through August because of this reduction in cooling capability . I would desire that the back flow preventer valves be completely removed, restoring the system to its original configuration. However, it is my understanding now that they are required by new federal regulations. So I would like to propose an alternate solution. I respectfully request that the 2-parallel lh inch back flow pre-venters be replaced with 2-parallel 3" or greater back flow preventers before May, 1980 in an attempt to alleviate this problem. I might add we noted a temperature increase when a back flow pre-venter was installed on Robeson Hall water supply earlier in the year. This was before back flow oreventers were installed on the individual supply of water to the re tor. This building back flow preventer pushed the temperature to just below our operating limit. When the other valves were installed in series with the building valve, it cou-pounded the problem and increased our temperature beyond the limit we

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I' Mr. James Clemens January 4, 1979 Page 2 e could satisfactorily operate. Your attention to this problem would be greatly appreciated. Sincer y yours, ,

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                                                  " Alan P. Curtner Reactor Supervisor APC/fbl cc:    T. F. Parkinson A. K. Furr A. H. Krebs J. B. Jones

p Attachment F 5  % e *w f f h'fM'-) I[ f Nd 3 'l

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2.5.1 Air Particulate Fission Product Monitor (APFPM) In the original reactor instrumentation, no provision was made for measuring any unplanned release of airborne fission products. t 'u r ren t l y , such instrumentation is needed to comply with both f ederal [2.2] and state [2.3] requirements. accordingly a new conitor system was installed. A block diagram of the system is shown in Figurt 2.9 An air sample is drawn from the reactor exhaust stack and passed through a moving filter tape. ihe activity on the filter tape is measured with a Na1 (Tl) scintillation crystal. inder normal operating conditions two short-lived fission products are detected: >.i-139 and Y-91: with gamma energies of 166 and 556 kev, respectively. hese fission products are pro-duced from traces of U-235 which are present as a contaminant on the reactor fuel plates. A plot of the gamma spectrum from the air particulates is shown in Fig. 2.10; these data were obtained with the MaI (TI) detector while the reactor operated at 100 KW. Pulses from the scintillation detector are ampli-fled and then routed to a single-channel analyzer (SCA), ratemeter, strip chart recorder and alarm circuit. The SCA window is calibratedeto accept those pulses associated with the photopeak of Ba-139. In the event of an abnormal release of fission products, the system output records the magni-tude of the release. Periodic calibration of the system with standard gacca sources allows an evaluation of any release of fission products f rom the exhaust stack. Af ter the gacna sensitivity of the detector is calibrated, the opera-tion of the entire system will be tested by irradiating a small sample of barium in a known flux for a pre-determined time. ine irradiated sample will then be placed on the moving tape and the iesponse of the APETM system will be measured.

I @ @ = @  ! I i i I L__________ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . , . _ _ ._ _ J U 000

1. FILTER PAPER DRIVE

2. NoI (TI) SCINTILLATOR, PHOTOMULTIPLIER AND PREAMPLIFIER (Horshow Model 858 Tennelec Model TCI55A)

3. HIGH VOLTAGE SUPPLY (Tennelec Model TC 951)

4. LINEAR AMPLIFIER AND SINGLE CHANNEL ANALYZER (Tennelec Model TC216)

5. RATEMETER, AUTORANGING (Tennelec Model TC595)

6. STillP CilART RECC3 DER (Rustrok Model 288) <

7. NIM OIN AND POWER SUPPLY (Tennelec Model TB-3/TC911)

Fig. 2.9 Schematic of the APFP Monitor e

Fig. 2.11) .\PFP fionitor Gamma Spec truin 1400 - l200 - m y POWER 100 kw E C O GAMMA SPECTRUM OF TAPE O O BLANK TAPE BACKGROUNC 1000 - o n 800 - 3 x x y

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