ML20237K572

Operations Annual Rept 1986-1987

ML20237K572
Person / Time
Site:	University of Missouri-Columbia
Issue date:	06/30/1987
From:	Brugger R, Meyer W MISSOURI, UNIV. OF, COLUMBIA, MO
To:	Berkow H Office of Nuclear Reactor Regulation
References
NUDOCS 8708190315
Download: ML20237K572 (82)

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h RESEARCH REACTOR FACILITY

i-i UNIVERSITY OF MISSOURI RESEARCH REACTOR FACILITY REACTOR OPERATIONS ANNUAL REPORT Ab%;i 1987 Compiled by the Reacter Staff

} Submitted b Walter A. Meye , Jr.

Acting Reactor Manager Reviewed and Approved 7 WD)M.BruggerLew Dr. Robert

/ f Director ,./ /

TABLE 10F CONTENTS Section Page Number

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'I. REACTOR OPERATIONS

SUMMARY

. . . . . . . . . . . . 1-1 through I-12 II. CHANGES TO THE STANDARD OPERATING PROCEDURES . . . . . . . . . . . . . . . . . . . . II-1 through 'II-5 REVISIONS TO THE HAZARDS

SUMMARY

REPORT ..... III-1 only III.-

IV. PLANT AND SYSTEM MODIFICATIONS . . . . . . . . . . IV-1 only V. NEW TESTS AND EXPERIMENTS ............ V-1 only VI. SPECIAL NUCLEAR MATERIAL ACTIVITIES ....... VI-1 through VI-2 VII. REACTOR PHYSICS ACTIVITIES . . . . . . . . . . . . VII-1 through VII-4 VIII.

SUMMARY

OF RADI0 ACTIVE EFFLUENTS RELEASED TO TiiE ENVIRONMENT . . . . . . . . . . . VIII-1 through VIII-2

'IX.

SUMMARY

OF ENVIRONMENTAL SURVEYS . . . . . . . . . IX-1 through IX-10 X.

SUMMARY

OF RADIATION EXPOSURES TO FACILITY STAFF, EXPERIMENTERS, AND VISITORS . . . . . . . . X-1 through X-2 2 4

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SECTION I ,

REACTOR OPERATIONS

SUMMARY

i 1 July 1986 through 30 June 1987 J

The following table and discussion summarize reactor operations in the period 1- July 1986 through 30 June 1987.

Full Power Percent

• Data Full Power Hours Megawatt Days of Total Time of Schedule July 1986 674.3 281.02 90.63 101.51 Aug. 198o 695.9 289.98 93.54 104.76 Sep. 1986 652.6 272.01 90.64 101.52

'Oct. 1986 685.9 285.85 ,92.19 103.25 Nov. 1986 646.4 269.35 89.78 100.55 Dec. 1986 632.4 263.59 85.00 95.20 Jan. 1987 698.1 290.91 93.83 105.09 Feb. 1987 632.8 264.24 94.17 105.47 Mar. 1987 711.6 296.50 95.65 107.12 Apr. 1987 647.8 269.96 89.97 100.77 May 1987 685.6 285.71 92.15 103.21 June 1987 629.4 262.25 87.42 97.91 Total for Year 7992.8 3331.37 91.24% of 102.19% of ,

time for yr. sched. time at 10MW for yr. at 10MW l

• MURR is scheduled to average at least 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> per week at 10MW.

Total time is the number of hours in a month or year.

1 1-1

JULY 1986 The reactor operated continuously in July with the following exceptions:

three shutdowns for refueling; two shutdowns for maintenance; and eleven unscheduled shutdown s.

Three nuclear instrument channel four (wide range monitor) high power rod run-ins (July 4,14,18) occurred as the regulating blade 60% witndrawn annunciation cleared during routine control blade shimming. No actual high power transients were observed on any instrumentation. This recurring problem was later traced to a faulty relay in the channel four RRI trip module.

On July 3, a nuclear instrument channel four high power scram occurred while manually lowering power from two kilowatts following a control blade reactivity worth procedure. This was caused by down-shif ting the channel four range switch too quickly during the power descent. The operator involved was instructed to more carefully monitor the instruments.

On July 4, a manual rod run-in was initiated when an operator detected a high voltage failure in the area radiation monitor. The high voltage tube was replaced and a hot startup was performed. As the reactor reached five megawatts, a reactor scram and isolation occurred when the area radiation monitor (ARM) again failed. All personnel exited the containment building as per proceduce for a Reactor Isolation. The shif t supervisor then re-entered the containment building eith a portable radiation monitor and found no unusual or elevated radiation read-ings. An electronics technician found and replaced a burned out resistor in the Area Radiation Monitor. The reactor was then refueled and returned to normal operation.

On July 11, a reactor scram was caused by a momentary electric power dip during a thunderstorm. A hot startup was completed and the reactor returned to normal operation.

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On July 14, a rod not in contact with magnet rod run-in occurred when the control blade "A" mechanism was bumped during a normal reflector sample handling

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evolution. The operator involved was instructed to be more careful and the reactor was refueled and returned to normal operation.

On July 24, a nuclear instrument channel number four (wide range monitor)

) high power rod ren *.1 occurred during a normal startup when the operator failed to shif t the wide range switch up-scale. During the subsequent startup, a channel four high power scram occurred when the same operator shif ted the wide range

, switch in the wrong direction. This recently licensed operator was instructed to I

concentrate more fully on the proper monitoring and manipulation of the wide range

, meter and switch.

On July 31, a channel four high power rod run-in occurred from less than 50 KW during a control blade reactivity worth procedure when the operator failed i

to shif t the channel four range switch. The operator involved was instructed to be more diligent.

Major maintenance items for July included: removing the iron collimator from beamport "F" and installing a depleted uranium filter; changing out control blade offse: "B"; replacing the high voltage tube and a resistor in the area radiation monitor; and installing the second 5-1/2 inch diameter sample wedge l 1

designated "N".

l AUGUST 1986 l

The reactor operated continuously in August with the following exceptions:

three shutdowns for refueling and/or flux trap sample changes; two shutdowns for maintenance and two unscheduled shutdowns.

On August 7, a reactor scram occurred while shif ting the wide range switch upscale during a normal reactor startup. The cause was suspected to be a sticking I-3

l relay contact in the feedback network of the picoammeter module for channel four.

The channel four range switch was cycled through its full range 'and the picoammeter relays operated normally. A normal startup was completed with no further problems.

On August 29, the reactor scrammed due to the loss of site power, verified by the University Power Plant. The reactor was refueled and returned to normal operation af ter electrical power was restored. l 1

Major maintenance items for August included: changing beamport "F" center- 1 I

tube filter from 40 cm to 25 cm of depleted uranium; dumping depleted pool DI bed j 1

"Z"; loading new pool DI bed "K" and replacing the pool flow scram relay. l g

SEPTEMBER 1986 The reactor operated continuously in September with the following exceptions:

four shutdowns for reiueling and/or flux trap changes; two shutdowns for mainte-nance; three shutdowns for shipping spent fuel; and two unscheduled shutdowns.

On September 18, a scram occurred when a lamp burned out in the light acti-vated switch section of the meter elay trip unit for pressure transmitter 944A (core discharge The lamp was rulaced and a hot startup was accomplished with no further pra ms.

On September 22, a nuclear instrument channel five high power scram occurred while the operators were physically adjusting the drywell for this instrument.

The physical adjustment of *he drywell occurred during a normal startup while reactor ower was being held at five megawatts as per Standard Operating Procedure.

The drywhil was inadvertently raised higher than anticipated during the adjustment, causing [ncreased response of the channel five detector and the high power scram.

Channels four and six indications and ecorders showed that no actual high power condition existed. The drywell was subsequently secured in an appropriate position and a n[rmal startup was completed with no further problems.

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An emergency preparedness appraisal was performed by two Region III Nuclear Regulatory Commission inspectors, J. P. Patterson and N. R. Williamsen, from September 8 through September 10, 1986. No violations of Nuclear Regulatory l

Commission requirements were identified during the course of the inspection.

l Major maintenance items for September included: shipping 24 spent fuel  ;

elements to Savannah River Plant, South Carolina; and replacing the lamp in the meter relay trip unit for pressure transmitter 944A.

OCTOBER 1986 The reactor operated continuously in October with the following exceptions:

two shutdowns for refueling; three shutdowns for maintenance; and one unscheduled shutdown.

On October 29, a nuclear instrument channel number four (wide range monitor) high power scram occurred while an electronics technician was working on the chan-nel four chart recorder. The scram was believed to be caused by electrical feed-back to the channel four drawer from the chart recorder, however the problem could not be reproduced while the electronics technicians were troubleshooting the chan-nel four drawer. No actual high power transient was observed on any instruments-tion and an operational test of all front panel nuclear instruments was satis-factory. The reactor was refueled and returned to normal operation.

Major maintenance items for October included: replacing the pool Th millivolt / current module; performing the scheduled biennial changeout of control blade offset "A"; and completing the scheduled visual inspection and gap verifi-cation of fuel element M0-200.

NOVEMBER 1986 The reactor operated continuously in November with the following exceptions:

two shutdowns for refueling and/or flux trap changes; two shutdowns for maintenance; and two unscheduled shutdowns.

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On November 7, two nuclear instrument channel number four high power rod run-ins occurred as the regulating blade 60% withdrawn annunciation cleared during routine shimming. In each case, no actual high power excursion was noted on any l instrumentation and af ter resetting the rod run-in, the reactor was promptly re- l l

turned to power. Electronics technicians isolated the problem to a faulty relay in j the WRM RRI trip module. l l

Major maintenance items for November included: replacing the #4 graphite l

wedge with a new graphite wedge #4C per special maintenance procedure; replacing the packing on tha liner of beamport "C"; and removing the boral plate on the front l l

of the Nuclepore irradiator case.

DECEMBER 1986 l l

The reactor operated continuously in December with the following exceptions:

four shutdowns for refueling and/or flux trap changes; three shutdowns for main +1-nance; and four unscheduled shutdowns.

On December 10, a reactor scram occurred from what was assumed to be a momen-tary site electrical power dip. No anomalies were identified and all plant systems were verified to be operating properly. The " white rat" scram circuit monitor was installed in both scram legs to help determine the cause of any future scrams of this nature. The reactor was refueled and a normal startup was performed with no further problems.

On December 19, a reactor scram occurred due to the failure of the scram trip light on the meter for core discharge pressure transmitter 9448. The meter was re-placed and the reactor was refueled and returned to normal operation.

On December 25, a manual rod run-in was immediately initiated by the duty operator when nuclear instrument charnel number five failed downscale from 105% to 75%. The detector for this instrument was replaced and the reactor was returned to normal operation.

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On December 31, a nuclear instrument channel number five high power rod run-in occurred soon af ter the completion of a ncirmal startup following a maintenance day. The indication for channel number five gradually increased while reactor and pool temperatures were in the process of stabilizing. This is a normal occurrence ehen primary and pool temperatures have been cooler than normal at startup. Normal procedure calls for reducing the gain potentiometer to an adequate level until the temperatures have stabilized. The duty operator failed to do this and the run-in cas initiated on channel number five at 114% power. The operator involved was instructed to be more alert to changing plant conditions. The rod run-in was quickly reset, the channel number five gain potentiometer lowered, and the reactor was returned to normal operation in approximately four minutes.

Major maintenance items for December includedt replacing the mechanical seal in primary pump 501A; dumping depleted pool D. I. bed "Y"; loading new pool D. I.

bed "R"; changing meters for pressure transmitters 928A, 944B, and 929; rebuilding antisip';on valves 543 A and B; and replacing N. I. channel number five detector .

(power range monitor).

JANUARY 1987 The reactor operated continuously in January with the following exceptions:

two shutdowns for flux trap changes; one shutdown for refueling; three shutdowns for maintenance and two unscheduled shutdowns.

On January 24, a nuclear instrument anomaly scram occurred when the detector for nuclear instrument channel number three (intermediate range monitor) failed. j The detector was replaced and the reactor was refueled and returned to normal operation.

On January 28, a rod not in contact with magnet rod run-in occurred when the housing of control blade "D" was bumped during a normal sample handling evolution.

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The auxiliary operator responsible was instructed to be more careful while per-forming this evolution in the future. The reactor was refueled and returned to normal operation.

Major maintenance items for January included: replacing the thrust and line bearings on pool pump 508B; modifying the personnel airlock door seal pressure gauges; rebuilding the solenoid air valves on the antisiphon sy;; tem valves; and replacing the' detector for N. I. channel number three.

FEBRUARY 1987, The reactor operated continuously in February with the following exceptions:

four shutdowns for refueling and/or flux trap changes; two shutdowns for mainte-nance; and one unscheduled shutdown.

On February 12, a tc ctor scram occurred as the wide range switch was down-shif ted while manually lowering reactor power af ter a low power run. The cause was suspected to be a sticking relay contact in the feedback network of the picoammeter module for channel number four. The switch was cycled through its normal range and the relays operated normally. A hot reactor startup was completed with no further problems. )

Major maintenance items for February included: replacing the detector for N.1. channel number six; replacing the DC amplifier in N. I, channel number five; and installing a new dodge coupling on cecondary pump number three.

MARCH 1987 The reactor operated continuously in March with the following exceptions: I i

three shutdowns for refueling and/or flux trap changes; three shutdowns for ma!n-tenance; one shutdown for a nuclear engineering class experiment; and one unsched-  !

l uled shutdown.

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On March 12, a channel number six (power range monitor) high power rod run-in occurred during'a reactor coolant temperature' coefficient experiment for a nuclear engineering class. The rod run-in was reset immediately and- the reactor was returned to normal operation in approximately one minute. The operator in-volved was instructed to be more diligent in monitoring available instruments.

On March 5,1987 during compliance testing of the reactor coolant high temperature scrams, meter relay unit 980A failed to initiate its required scram in response to a simulated high temperature. The details of this instrument failure were reported in a Licensee Event Report dated March 31, 1987.

Major maintenance items for March included: replacing the reactor coolant inlet temperature meter relay (scram) trip unit 980A; pulling back beamport "F" centertube approximately four feet; and replacing the detector for nuclear instru-ment channel number two (intermediate range monitor).

APRIL 1987 The reactor operated continuously in April with the following exceptions:

three shutdowns for refueling and flux trap changes; three shutdowns for main-tenance; two unscheduled shutdowns; and one shutdown because of high airborne particulate level in the containment building.

On April 10, a rod run-in occurred with no annunciator alarms or other .

indications. Af ter the duty operator determined that systems were functioning )

normally, the rod run-in was reset and the reactor was returned to normal opera- ')

tion in about two minutes. I On April 15, the reactor scrammed due to the loss of site electrical power.

The emergency generator satisfactorily assumed the electrical load for necessary equipment. Site power was restored in approximately four hours and the reactor was refueled and returned to n

• mal operation.

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On April 21, a manual reactor isolation was initiated due to an unusually high (radioactive) airborne particulate level in the containment building. All personnel exited the containment building as per the reactor isolation procedure and remained outside until health physicists determined that the radionuclides levels were less than one-tenth MPC limits. The containment building was then re-entered and the reactor was refueled and returned to normal operation. The cause of the high particulate activity was determined to be the performance of a main-tenance procedure for the Nuclepore experiment without first venting the irradi- )

k ator case. j On P.pril 3, an emergency preparedness drill, simulating high airborne I l

activity in containment, was satisfactorily completed.  !

Major maintenance items for April included: replacing the scram trip units in the meters for primary pressure transmitters 944 A/B and 943; replacing the motor bearings in the main air compressor; and completing the biennial changeout of control blade offset "D". j MAY 1987 i

The reactor operated continuously in May with the following exceptions.  !

one shutdown for refueling; two shutdowns for maintenance; and two unscheduled s hutdown s.

j On May 6, a rod not in contact with magnet rod run-in occurred due to a momentary electrical power interruption which was verified with the power plant.

The reactor was subsequently refueled and returned to normal operation.

On May 8, a manual rod run-in was initiated by the duty operator upon the failure of the area radiation monitoring system (ARMS). The high and low voltage power supplies for the ARMS were repaired and the reactor was refueled and re- 3

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turned to normal operation.

1-10 1 1

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l4ajor maintenance items for liay included: replacing the detector for nuc-lear instrument channel number five (power range monitor); replacing the detector for nuclear instrument channel number three (intermediate range monitor); re-pairing the power supplies on the area radiation monitoring system; and com-pleting a satisfactory annual containment building leak rate test.

JUNE 1987 The reactor operated continuously in June with the following exceptions:

two shutdowns for refueling; two shutdowns for maintenance; four unscheduled s hu tdown s.

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On June 11, an anti-siphon line high level rod run-in occurred withir Tive minutes af ter a normal startup. The anti-siphon line was quickly drained and the rod run-in was reset. The reactor was returned to normal power in approxi-mately four minutes. The anti-siphon system had not been adequately blown down prior to the startup and the operator responsible was instructed to be more diligent in performing this task.

On June 18, a reactor scram occurred while the reactor was subcritical at l less than 100 KW. At the time, the reactor was in the process of being shut down and a control blade drop time compliance check was being performed. No l l

specific cause for the scram could be determined. The " white rat" scram circuit monitoring system was installed to help detect any future occurrences of this ty pe. I Late on June 18, a reactor scram occurred again while the reactor was subcritical and a blade drop time compliance check was being performed. This scram was initiated by pressure transmitter 944 A/B (core discharge) and was attributed to the fact that the mercoid switch for pressure sensor 941 (which regulates nitrogen pressure in the pressurizer) had been adjusted slightly from I-11

its desired setpoint during maintenance earlier in the day. The switch was re-adjusted and the reactor was returned to normal operation.

On June 20, a reactor scram occurred with reactor loop low flow indicated on the annunciator. A check of the digital and chart recordings indicated that no actual low flow condition existed at this time. A spurious signal from a reactor loop flow transmitter was suspected and the transmitters were subse-quently vented. A hot startup was then completed with no further problems of this type.

Major maintenance items for June included: replacing a primary pump and i

impeller (501A); replacing the mercoid switch for pressure sensor 941 (nitrogen  ;

. system); and replacing the sealing gaskets for motor doors 504 and 505.

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SECTION II CHANGES TO THE STANDARD OPERATING PROCEDURES Revised October 1981 1 July 1986 through 30 June 1987 As required by the MURR Technical Specifications, the Acting Reactor Manager reviewed and approved the changes to the Standard Operating Procedures and the Emergency Procedures.

A. STANDARD OPERATING PROCEDURES:

There have been 2 revisions (22 and 23) made to the Revised October 1981 l

manual during the year. l 1

B. EMERGENCY PROCEDURES: )

l As required by the MURR Technical Specifications, the Acting Reactor l l

Manager reviewed and approved the changes to the Emergency Procedures.

There was 1 revision (#5) made to the Site Emergency Procedure Section ]

of the Emergency Procedures and 1 revision (#3) to the Facility Emergency Procedure Section of the Emergency Procedures.  ;

l The revisions to the Standard Operating Procedures and Emergency Procedures are contained in this section with the part of each page that was revised marked on the right side of the page by a bracket (3). <

l 11-1 I

J REVISION NUMBER 22 TO OCTOBER 1981 MANUAL l

Page Number Date Revised I

S0P/I-2 10/8/86 l i

SOP /I-9 10/8/86 l S0P/I-13a 10/8/86 S0P/I-13b 10/8/86 S0P/ I-13c 10/8/86 SOP /11-1 10/8/86 SOP /VI-9 10/8/86 SOP /VI-10 10/8/86 SOP /VIII-3 10/8/86 SOP /VIII-4 10/8/86 S0P/A-la 10/8/86 SOP /A-8a 10/8/86 SOP /A-8b 10/8/86 SOP / A-10a 10/8/86 SOP /A-10b 10/8/86 S0P/A-11-a 10/8/86 II-2

1.4 Administrative Policies 1.4 .1 Standing Orders New procedures prior to.becoming part of these Standing Operating Pro-cedures, procedures to be in effect for a short period, and special instructions relating to operation of the reactor, will be issued as Standing Orders. All effective Standing Orders, as well as a listing of current orders, will be maintained in a log in the reactor control room. . . Standing Orders that are no longer applicable or that are incor-porated into the-Standard Operating Procedures will be cancelled and re-moved from the log. All Standing Orders will be approved by signature of the Reactor Manager or his authorized delegate.

1.4.2 General Operating Policies 1

A. Safety Safe operation of the reactor will take precedence over other considerations.

B. Supervisory Authority The importance of one coordinator for all reactor activities is recognized for safety and effective control of operation. The Reactor Manager will have complete operating authority over all activities related to reactor operation.

C. Operating Crews The minimum number of reactor operators for reactor operation will be two licensed persons. One of these will be licensed as a senior I operator. There will be one licensed operator in the control room at all times whenever the reactor is not considered secured, as de-fined in Part 1.1 of the Technical Specifications. Exception: It is not necessary to have the control room manned during refueling.

These activities may be directed from the reactor bridge.

D. Console Log A log will be maintained in the control room by the reactor operator providing a detailed diary of reactor operation. Any corrections to ] ,

a log entry will be done by a single line through the incorrect ]

entry, plus the initials of the person making the change and the ]  !

date. ]

Rev. 10/8/86 App'd luMV\'

SOP /I-2 l

operation should be determined. The primary power calculation will normally be used to establish the 10 MW power level, however, the nominal steady state power level shall not exceed 10 MW. The reactor shall not be operated at a power level which causes the steady state secondary calorimetric to exceed 10.5 MW unless it is confirmed that the secondary calorimetric is in error or out of commission.

The primary system DI flow bypasses the core and yet it flows through the primary flow orifice. Therefore, the primary flow as read on the recorder should be decreased by the primary 01 flow before the value is used in calculating the power level.

The recorder values should be logged on the log sheets without correction.

I. Steady state reactor powers of 1 MW and greater will be deter-mined by the method stated above. The power indicated by Channels 4, 5, and 6 shall be maintained greater than 100% during ]

steady state full power operation. Channels 4, 5, and 6 are adjusted by proper positioning of the drawer amplifier feedback '

potentiometers. After adjustment of a potentiometer, the change l in indicated power shall be logged in the console log and the new pot setting logged on the Startup Nuclear Data Sheet. The Shift Supervisor's approval must be obtained before adjustmen+

of any Power Range Monitor. Adjustments shall only be made after a determination of the power level by heat balance.

J. Minimum nuclear instrumentation for normal operation shall be two (2) intermediate range channels with period trip, two (2) power range channels each with high flux trips and one (1) wide range channel with high flux trip.

K. During reactor operation the control room shall be occupied only by persons authorized by the Reactor Manager or Shift Supervisor.

Rev. 10/8/86 App'd bl&7Yi S0p/I-9

I 4.9- Physical Protection of Special Nuclear Materials in accordance with 10 CFR 73, special requirement s must be met in safeguarding Special Nuclear Material. The safeguards provided and the procedures applicable to maintaining the security of Special Nuclear Materials are contained in the f acility Security Plan and Security Procedures.

1.4.10 Equipment Tag Out Procedure I.4.10.1 Purpose The purpose of tagging equipment is to prevent damage to equipment or personnel or to provide supplemental information concerning a special operating procedure.

I.4.10.2 Tagging Equipment Three tags will be used. See Appendix 4 for a sketch of the A.

danger tags. The tag color designates its purpose as follows:

Red - Danger to personnel Yellow - Danger to equipment White -Supplemental information concerning an operating procedure B. The Shift Supervisor's approval must be obtained before equipment is tagged.

C. All tags and the tag log must be filled out and tags attached to equipment by a licensed operator. The tag log must contain a description of the equipment tagged, reason for tagging, date, and initials of operator initiating the tag.

1.4.10.3 Removing Tags A. The Shift Supervisor's approval must be obtained before any tags are removed.

B. When the tags are removed,, the date of removal and the initials ,

of the individual clearing the tags shall be entered on the 4

tag log.

l Rev.10/8/86 App'd Q&f4 SOP /1-13a ]

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C. Ensure the valve or equipment from which the tag is removed ]

is placed in the proper position or condition. ]

I.4.10.4 MISSING TAGS ]

A. Immediately notify the shif t supervisor in the event a tag ]

is determined to be missing. ]

B. The shift supervisor will investigate the cause of the ]

missing tag and will replace it with a new tag if necessary. ]

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Rev.10/8/86 App'd iL @ S0P/I-13c ]

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l i Section II REACTOR OPERATING PROCEDURES l

11.1 Routine Reactor Operation ,

11.1.1 Procedure for Reactor Startup For full power, closed pressure vessel operation, the reactor will ]

be brought to its scheduled operating power level according to the procedure outlined below. l NOTE: Starting a secondary pump during a normal reactor startup ]

may reduce primary temperature and pressure to the' point ]

that core discharge pressure transmitters 944 A/R may ]

scram the reactor. ]

A. Take a complete set of full power process data.

B. Obtain from the Shift Supervisor an estimate of the critical bdnked Cortrnl blade position.

C. Take a complete set of nuclear data on the Startup Nuclear Data Sheet.

D. Complete the applicable star" - -hacksheet required by Section I (1.4.3.F).

E. Obtain from the Shift spr v ,r pei 'issic. to commence a reactor startup.

F. Announce via the publi <

ystem that a normal reactor startup has been commen, G. Withdraw the four control ! et s in gang, stopping to take a set of startup nuclear di a P' five-inch increments. Indicate in the console log book that startup has commenced.

H. When the blades have reached a position within 2 inches of the estimated critical position, discontinue pulling in gang and take a set of startup nuclear data.

I. Continue the startup, withdrawing only one blade at a time until the reactor power level is increasing on noo less than a 30-second period.

J. At a point where the reactor is indeed critical and on a positive period, a console log entry shall be made stating that fact.

Rev. 10/8/86 App'd M SOP /II-1

I VI.6.3 Secondary Water pH Control To control pH, water is sampled and monitored by the pH unit located in the tunnel entrance of room 114. If the pH increases above the system setpoint, the acid injection valves automatically open and acid is gravity fed into the tower sump. The acid used is concentrated sulfuric acid supplied from the 250-gallon day tank in the cooling tower. See Section 11.4 for specifications.

VI.6.4 Sample Paths for pH and Conductivity A. During normal operation of the secondary system, the automatic l pH and conductivity control units receive their sample water j through valves S104 and S151. j B. During operation of the secondary system with the air condi-tioning units secured and isolated, close S104 and open S103.

This provides a representative sample for these units to control pH and conductivity. To shift the automatic blowdown system during operation with the air conditioning units isolated, close S102 and S101.

C. During shutdown periods, either secure the pH and conductivity control units or run at least one large secondary system pump.

Sample response time from the cooling tower basin is much too slow with only P-4 running.

VI.6.5 Secondary Water Corrosion Prevention A. The prevention of corrosion in the secondary system is accomplished by automatic addition of a corrosion inhibiter.

B. These chemicals are fed automatically by a metering pump ]

system based on makeup flow. ]

Rev.10/8/86 App'd 1)$W\ SOP /VI-9 i

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c VI.6.6 Secondary Silt, Algae and Mud Control Silt and mud buildup is controlled by the feeding of a chemical ]

silt dispersant to the cooling tower basin. The dispersant is ]

added to ensure solids remain suspended a sufficient amount of ]

time to allow the secondary blowdown to remove them from the system. This. reduces secondary conductivity and minimizes the buildup of silt in low flow areas, a fouling condition.

Microbiological and algae growth is controlled by the addi-tion of two microbiocides/algecioes. The addition frequency is determined by weather conditions and reactor operations. ]

i Rev. 10/8/86 App'd BMkn 50P/VI-10 i I

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This might be done because of unusual hazards, special

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conditions involved or because the RSSC feels that an unreviewed safety question does or may exist.

5. The Reactor Advisory Committee (RAC) will normally review an experiment only if the experiment has been referred to it by the RSSC. If the RAC determines that the experiment does not involve an unreviewed safety question the review process is complete.

6. If the RSSC and/or the RAC feel that a proposed experi-ment introduces an unreviewed safety question the experiment must be submitted to the NRC for final revi ew. The MURR staff will generally prepare the documents necessary for submittal to the NRC.

C. Approval of RUR After all of the reviews have been completed the Reactor Manager will indicate on the RUR any additional limitations required beyond those listed in the data package and will then sign the RUR as being approved. Copies of the approved RUR will be distributed to:

1. The experimenter

2. Facility Director

3. Manager of Health Physics ,

4 Reactor Safety Subcommittee

5. MURR Staff

6. Reactor Advisory Committee (if RAC was involved in the l

review) j

7. Reactor Service Engineer D. RUR Review and Update ]

1. The RUR's will be reviewed by the Reactor Manager and ]

the Principal Experimenter on an annual basis. ]

2. The Principal Experimenter may use the RUR review to up- ]

date the experiment description; to request changes to 3 l

the experiment; to change the list of authorized users ]

l on the experiment; and to familiarize himself with the ]

limitations placed on the experiment. ]

3. The Reactor Manager will use the RUR review to ensure ] l that the experiment descriptions, activities, isotopes, ]

handling procedures, license considerations and safety ]

analysis are valid for the current range of experiments. ]

Rev.10/8/86 App'd M SOP /VIII-3 1

VIII.1.2- Flammable or Toxic Materials A. Definitions

1. Flashpoint of the-liquid shall mean the temperature at which it gives off vapor sufficient to form an ignitable mixture with the air near the surface of the liquid or within the vessel used as determined by appropriate test procedures referenced in Section 1910.106 of the OSHA Regulations.

2. Liquid shall mean, for the purpose of this section, any material which has a fluidity greater than that of 300 penetration asphalt when tested in accordance with ASTM Test for Penetration for Bituminous Materials, D-5-65.

3. Combustible liquids shall mean any liquid having a flashpoint at or above 140 F (60 C).

4. Flammable liquids shall mean any liquid having a flashpoint below 140 F and having a vapor pressure not exceeding 40 pounds per square inch (absolute) at 100*F.

a. Class I liquids shall include those having flash-points below 100 F and may be subdivided.

(1) Class IA shall include those having flash-points below 73 F and having a boiling point l below 100 F 1-(2) Class IB shall include those having flash-l points below 73*F and having a boiling point at or above 100'F.

(3) Class IC shall include those having flash-points at or above 73*F and below 100 F.

b. Class II liquids shall include those having flash-points at or above 100"F and below 140 F.

c. Class III liquids include all combustible liquids.

5. Safety can shall mean an approved container, of not more than 5 gallons capacity, having a spring-closing lid and spout cover and so designated that it will safely relieve internal pressure when subjected to fire exposure.

B. Limitations of Flammable and Combustible Liquids

1. No flammable or combustible liquids may be taken into  !

the containment building unless they meet the require-ments of Table VIII-1 below.

@m 10/8/86 A3@'d % _ SOP /VIII-4

REACTOR STARTUP CHECKSHEET DATE:

FULL POWER OPERATION (or Low Power Forced Circulation) TIME (Started):

BUILDING AND MECHANICAL E001PMENT CHECK' LIST l l 1. Run emergency generator f or 30 minutes and check the governor oil level.

(Required if shutdown for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or af ter each maintenance day.)

2. a. Check operation of f an f ailure buzzer and warning light. -

3 (Required if shutdown longer than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.)

b. Test stack monitor and low flow alarm per SOP while in west tower. ]

3. Visual check of room 114 eauipment completed.

a. P501A and P501B coolant water valves open.

b. 51 and S2 hydraulic pumps on (oil level normal).

c. Pump controllers unlocked to start (as required). I

d. Insure N2 backup system on per SOP.

e. Open air valve for valve operating header (V0P 31),

f. N2 backup valve open.

g. Check valves 599A and'599B open.

h. Pipe trench free of water (after maintenance day, check the four-pipe annulus drain valves for water leakage).

i. Add DI water to beamport and pool overflow loop-seals.

j. Vent the 6000 gallon pool hold up tank.

4. Visual check of CT and secondary equipment completed.

a. Oil level in CT fans normal (after maintenance day).

b. Secondary makeup isolation valve power switch closed.

5. Beamport Floor

a. Beamport radiation shielding (as required).

b. Unused beamports checked flooded (after maintenance day).

c. Seal trench low level alarm tested (after maintenance day).

6. Emergency air compressor (load test for 30 minutes after maintenance day).

7. Emergency Pool Fill (check valves PIV-1 and PIV-2 locked open), )

8. Reactor Pool

a. Reflector experimental loadings verified and secured for start-up.

b. Flux trap experimental loading verified and secured for start-up, or strainer in place.

l l c. Check power on and reset, as necessary, silicon integrator, totalizer setting, silicon rotator and alarm system.

REACTOR CONTROL SYSTEM CHECKLIST

1. All chart drives on; charts timed and dated. IRM recorder to slow.

2. Fan f ailure warning system cleared.

3. Annunciator board energized; horn off.

4 Television receiver on.

5. Primary / pool drain collection system in service per S0P.

6. Secondary system on line per SDP (as needed).

7. Primary system on line per S0P.

I I a. Primary cleanup system on line.

8. Pool system on line per SOP.

a. . Pool cleanup system on line.

b. Pool skimmer system vented.

c. Pool reflector aP trips set per 50P.

9. Valves S1 and S2 cycled in manual mode and positioned as required.

10. Nuclear Instrumentation check completed per S0P. 1

a. The following trip values were obtained during the check:

IRM-2, run-i n seconds (11+1) Scram seconds (9+1) 1RM-3, run-in seconds (1171) Scram seconds (971)

WRM-4, run-in  % (114+1) - Scram  % (119+1) ~

PRM-5, run-i n  % (11471) Scram  % (119T1)

PRM-6, run-i n  %(11411) Scram  % (11971)

11. Channel 4, 5, and 6 pots returned to last heat balance position.

12. SRM-1 detector response checked and set to indicate > 1 cps.

Rev. 10/8/86 App'd brM SOP /A-la

WZmCTDM WDU13WE FMYWOL De8@

Time of start of patrol.

Time and date all charts.

Check ARMS trip settings. ,

Visual check of entire pool.

Anti-siphon tank pressure. 36 psig i 3 psi North iso door seal press. 18-28 psig

~~

South iso door seal press. 18-28 psig I~~5th level backup doors. Open I~~5th level detector reading. 0-3.5 mr/hr I~~5th level trip point set. 3.5 mr/hr

> 16" iso viv A air pressure. 45-55 psig 4

[~~E5ergcompressonstandby. c v open,

((heggs{g

> Containment hot sump pumps. Operable -

l 3 Door 101 seal pressure. 18-28 psig 3 8P Floor Conditions normal.

3 Fuel Vault Locked 3 Inneraiflockdoorsealpress. 18-28 psig a Outer airlock door seal press. 18-28 psig

=> l 0 T-300 level. > 2000 gal. l 1

0 T-301 level. < 6000 gal. '

o Labyrinth Sump Level < Alarm Pt.

o RO Unit ON Power  ;

a RO Unit Temp / Pressure 25-28 C/190-200 psig/ ]'

standby Thermostat > 50"F a

LG Rm.(EG(Batt.

OP check Wed.

switch to Automids.);

Temp > 40*F (Gas I sight glars )

o T-300, 301 Room Thermostat > 55*F l Thermostat > 40 F

. Rn. 114 Particulate filter aP < 2.5" H 2O l the first routine patrol of the day or the first patrol af ter a startup, drain all water ,

'om the anti-siphon system. If draining causes the pressure to drop significantly, return J

> tha middle of the band (36 psig) and record the pressure here. If a condition or reading i normal, enter a "/" (for conditions) or the reading in the applicable box. If the con-tion is abnormal, enter the condition or reading and circle it. Explain all abnormal con-tions or readings in the remarks.

l MARKS:

rv. 10/8/Fb App'd lu7tA SOP /A-8a

External Doors 'fhen!$$n$$h l ]

U

)

CT basin water level. 5-10" ]:

Automatic Seconoary Makeup Valve'Open/ Auto ($ ating) 3

~

Acid day tank level. Visible )

l CT sump pumps. Operable P-pump (s) running.

Pump strainer $. 0-7.0 psi Discharge pressure.

Pump strainer $. -

0-7.0 psi Discharge pressure.

Tunnel sump pumps. Operable WT booster fan. Running Acid control and pH. Flow pH 7.8400-g00cc/

- .2 min

, Blowdown control /cond. g,5gog,0g/ min

> Fission Product Monitor Flow 95-105 cc/ min Viv control header pressure.90-120 psig .

> Pressurizer N2 supply press.90-100 psig ,

[ Check Rm 114 from door.

3 Deltech oil filter " red level" and blowdown.

< 75% dark red 3 Seal trench. {n 6 3 Full Np bottles. Total > 3 o Bank A bottle press. > 250 psig o Bank B bottle press. > 250 psig o Bank on service. A or B .

c N header pressure. 135-145 psi 2

o Waste tank #3 level.

. Waste tank #2 level. I

. Waste tank #1 level. j

.. Doors to Ct, WT's, Demin. Locked Rm. 114 and CT Tunnel.

i. Time of Completion of Patrol.

'. Operator Initials i I

MARKS

. )

9% 10/8/ud AB@' d h SOP /A-Bb l

~~ ~

RADIATION WORK PERMIT RWP No.' Date:

Time: -1.ocation:

Work

Description:

N RWP expires 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> f rom above .

date and time unless extended to: DATE .7J ME SIGNED

\ ' PROTECT 4VE EOUIPMENT REQUIRED:

RADI ATION SURVEY RESULTS: ,

General Area: 'tnrem/hr _S.TI. Dos. _ Finger Rings

, ~

Hot Areas: _ Lab Coats - _ Coveralls

.f4*

mrem /hr at 43 loves: Latex f . . . . . _ Rubber s.m .

rnrem/hr at /. -4 7.

j . { _;gOs Shoe Covers: .. _ Nylo_npJ3ubber...

Boots wremthr at >

-J

_w

y.

-r.,

g.:

- e- ;L~

R i g gs  %'$.f m ,.

CONTAMIN ATiON SU RVEY RESULTS: Efg p. Equipmentq ; kj,;, g,,_ ,,;,7y,sn -

.. e g+:> ;= .. :

i iiFace . _.

4'3>E"" x. 0?%i,ll FM. . .: .

General Area: - dpm/100 cm2 . _ . . . _.. _ . .

s

. , w.$ ; := ,e.up91;f

~2 Other

~

Other ' Areas: (v _ . . . . .' a : T

,,.-g..q qi: . .g;

- dpm/100 cm2 \ '- ?$qigiMnN'

n a.~n..  ;,= G.

\

- dpm/100 cm2 ,

\

. .-g,'y nWp i \ . ~~.j:7

,. u. . r. .

. .g33 : -

Continuous H.P. Monitoring intermittent H,.P.?Aoriitoring REQUIRED PRECAUTIONS:

Tool & Equipment Survey Rope OfI and Post Area H & F Count On Leaving Job i ,

Personal Survey on Leaving Job Air Monitoring Special Time Dose Lim!

! opef 811Dn $hiti Supervisor Read F oi information ,

begned o8lf Date Time APPROVALS:

l H.P. Summary Report? Yes No i

Hwnh bics Date "f eme g, ,

if yes, attach copy to this RWP.

a. TM80@% na nn n. ,aa n _

3CI AL INSTRUCTIONS:

RSONNE L INVOLVED

Total Total Est. Est.

Dose Dose 0 mrem 13) mrem g mrem 14) mrem g mrem 15) mrem

) mrem 16) mrem

5) _ mrem 17) mrem  ;

3) mrem 18) mRam

?) _ mrem 19) mrem

)) mrem 20) mrem

)) mrem 21) mrem

)) mrem 22)__ _ mrem l} mrem 23) mrem

?) mrem 24) mrem s

9. COMMENTS gned Date m am m or.w n__nm nn -

No.

~

ETURN 0RIGINAL TO HEALTH PHYSICS OFFICE WASTE TANK SAMPLE REPORT

'ANK NO. TANK LEVEL (Liters)

Emplcted adding water to this tank. TIME DATE AMPLER TIME DATE

1. Analysis Results Nuclide Half Life Physical Form Concentration MPC Activity (oci/ml) (uci) a H.3 12.3Y

b. .

1 pH FRACTION OF MPC ]) i Analysis by TIME DATE )

l Concentration (tci/m) Total Volume (liters) Activity (mci) j x =

(a) x =

(b) f

2. Approvals Required for: j Any Discharge ...................... '

l Shift Supervisor '

Discharge of > 10 mci of H 3 , > 2 mci of other activity, or ,

to Secondary System . .................

1 Reactor Manager Discharge Limit Approved . . . . . . . . . . . . . . . . .

Health Physics

3. Action Taken Data Discharged Time Discharged Volume Discharged (Liters)

Tank Discharged to (check one) Sanitary Sewer Secondary System Not Discharged REMARKS Rev.10/8/86 App'd hum SOP /A-lla

4 REVISION NUMBER 23 d

T0 OCTOBER 1981 MANUAL Page Number Date Revised S0P/I-3 12/16/86 S0P/I-4 12/16/86 S0P/I-12 12/16/86 SOP / II-7 12/16/86 SOP /A-la 12/16/86 SOP /A-lb 12/16/86 SOP /A-6b 12/16/86 1

P e

o Y

11-3

E. Changing Reactor Reactivity.

Operations affecting changes in core reactivity, other than  ;

normal steady state power control, i.e. refueling, startup, etc., will be directly supervised by a senior operator.

F. Malfunction of Reactor Systems While at Power Any malfunction or abnormal cperation of a control or reactor system should be immediately brought to the attention of the Shift Supervisor. The decision as to whether to continue operation of the reactor depends upon the severity of the malfunction. It remains with the Shift Supervisor as to what ,

immediate action need be taken. However, the duty operator is )

authorized, in the absence of the Shift Supervisor from the control room, to place the reactor in a safe shutdown mode if he deems it necessary. It is important that the reactor systems, while the reactor is critical, not be experimented 1 l

with, or reactor control systems tested, unless permission to j do so is explicitly granted by either the Reactor Manager or i his designated representative. l G. Maintenance Performed on Reactor Systems ] l All maintenance on reactor and license related systems and ]

equipment will be reviewed by the shift supervisor or acting ]

shift supervisor to ensure operability of these systems prior ] I to reactor operation. ]

H. Reactor Operating Parameters The reactor shall be operated in strict accordance with the -  !

Reactor License R-103 and the operating limits in Tables III and IV.

I 4.3 Startup A. Startup Following a Scram The reactor will not be started up following a scram until the cause of the scram has been determined and safe corrective action taken. If, after thorough investigation, the cause of the unscheduled reduction in power cannot be determined and all systems are found to be normal, the reactor may be started up with the approval of the Reactor Manager or his authorized delegate.

Rev. 12/16/86 App'd hM SOP /I-3

B. Occupancy of the Reactor During reactor startup or transient operation, occupancy of the control room and containment building will be limited to the reactor staff, experimenters and those observers as approved by the Reactor Manager or duty shift supervisor.

C. Operator Change During Transients Control of the reactor will not be transferred from one operator to another during power transient operations.

D. Control Blade Operation

1. The control blades shall not be moved in gang control

. after the reactor is critical except to reduce power (11.1.3), shutdown the reactor, or as part of the automatic shimming operation.

2. The ganged control blades, or any single control blade, will not be withdrawn simultaneously with the regulating blade.

E. Hot Startup A startup within two hours of any shntdown, in which re-start l capability is in doubt, shall be called a hot startup. A I hot startup shall only be made by the Shift Supervisor, senior

)

reactor operator, or a licensed reactor operator under the direct and close supervision of a senior reactor operator.

The approach to critical may be made entirely in gang control of the control rod drives.

F. Startup Checksheet

1. A Startup Checksheet will be completed before nuclear 4 operation of the reactor. Separate Startup Checksheets will be used for low and full power operation. Operation I above 50 KW will require completion of the full power checksheet. Operation at or below 50 KW with forced convection cooling will also require completion of the Full Power Checksheet. The completed checksheets will be approved in writing by the shift supervisor or acting ]

shift supervisor before withdrawing the control rods for ]

the startup. A short Rev. 12/16/86 App'd WDfA SOP /I-4

l B. After the supervisor has completed the top portion of the RWP, he will sign and date the form and deliver it to Health Physics, who will number the RWP, conduct the necessary surveys and determine the protective measures necessary t~or the job. Health Physics will then complete the remainder of the RWP indicating the survey results and the protective measures required. He will then sign and date the RWP and return it to the supervisor.

C. The 5,upervisor will post the approved RWP at the job site for ready reference by the personnel doing the work. He will removg the tear-off strip and hang it up on the hook in the control room.

D. When the job has been completed and the job site has been ]

cleaned up and decontaminated, the supervisor will sign off the RWP and deliver it to Health Physics. After Health Physics has verified that the job site is clean and decon-taminated, he will sign off the RWP and maintain it in his RWP file. The supervisor will remove and discard the tear-off strip from the hook in the control room.

I.4.8 Radiation Safety i

The Shift Supervisor is directly responsible for the overall safety of personnel on his shift and indirectly responsible for all personnel whose safety may be affected by activities i conducted under his supervision. Radiation safety is a very important part of this responsibility. It should not be construed that surveys, monitoring, or other measurements to check for contamination or radiation are to be made by operations personnel, but rather that the Shift Supervisor is responsible to insure that through coordination with the Health Physics personnel, adtquate protection is provided for evolutions conducted during his shift.

Rev. 12/16/86 App'd hk SOP /I-12 1

the desired power level, each control blade must be withdrawn individually if its position .is within 2.00" of -

l its previous critical position. Maintain all blade posi-tions such that the difference in position between any two blades is not greater than 1.00".

D. Recover criticality at the new power level.

E. Replace the reactor control in the automatic mode, if desired.

F. Take a complete set of nuclear data.

G. Switch the IRM recorder from fast to slow speed.

H. Record the power change in the log book and on the Startup Nuclear Data Sheet as required.

For power level. maneuvers at power less than 1 MW:

A. For increasing or decreasing power, the same procedure as outlined in parts A-H above apply, with the exception that power level increases up to 1 NW may be made on periodsin_o o less than 30 seconds.

11.1.5 Procedure for Control Blade Shimming A. The Automatic Shimming Unit can be considered operational and will shim the regulating blade out should it reach 5.20", but it should not be used for routine shimming.

B. For manual shimming in auto control, each control blade will be withdrawn or inserted individually while the regulating blade automatically adjusts to maintain a constant power level.

C. The control blade positions will be adjusted such that they all read the same position across the board after the shimming operation has been completed.

D. During the shimming operation, the Reactor Operator will closely observe the nuclear instrumentation and the movement of the regulating control blade (when in automatic control), ]

to verify that the system response to the shimming operation is normal.

i l

l Rev. 12/16/86 App'd dDM SOP /II-7 1

REACTOR STARTUP'CHECKSHEET DATE:

FULL POWER OPERATION (or Low Power Forced Circulation) TlHE (Started):

BUILDING AND MECHANICAL EOUIPMENT CHECKLIST l l 1. Run emergency generator for 30 minutes and check the governor oil level. j (Requi. red if shutdown for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or af ter each maintenance day.)

l l 2. a. Check operation of fan f ailure buzzer and warning light.

(Required if shutdowr langer than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.)

l l b. Test stack monitor and low flow alarm per SDP while in west tower.

3. Visual check of room 114 equipment completed.

a. P501A and P501B coolant water valves open.

b. 51 and S2 hydraulic pumps on (oil level normal).

c. Pump controllers unlocked to start (as required).

d. Insure N2 backup system on per SOP.

e. Open air valve for valve operating header (V0P 31).

f. N2 backup valve open.

g. Check valves 599A and 599B open.

h. Pipe trench free of water (after maintenance day, check the four-pipe annulus drain valves for water leakage),

i. Add D1 water to beamport and pool overflow loop-seals.

j. Vent the 6000 gallon pool hold up tank.

4. Visual check of CT and secondary equipment completed,

a. Oil level in CT fans normal (after maintenance day).

b. Secondary makeup isolation valve power switch closed.

5. Beamport Floor

a. Beamport radiation shielding (as required).

b. Unused beamports checked flooded (af ter maintenance day).

c. Seal trench low level alarm tested (after maintenance day).

6. Emergency air compressor (load test for 30 minutes after maintenance day).

7. Emergency Pool Fill (check valves PIV-1 and PIV-2 locked open).

8. Reactor Pool

a. Reflector experimental loadings verified and secured for start-up.

b. Flux trap experimental loading verified and secured for start-up, or strainer in place.

c. Check power on and reset, as necessary, silicon integrator, totalizer l l setting, silicon rotator and alarm system.

REACTOR CONTROL SYSTEM CHECKLIST

1. All chart drives on; charts timed and dated. IRM recorder to slow.

2. Fan failure warning system cleared.

3. Annunciator board energized; horn off.

4. Television receiver on.

5. Primary / pool drain collection system in service per S0P.

6. Secondary system on line per S0P (as needed).

7. Primary system on line per 30P.

a. Primary cleanup system on line.

8. Pool system on line per SOP.

a. Pool cleanup system on line,

b. Pool skimmer system vented.

c. Pool reflector oP trips set per 50P. '

9. Valves S1 and S2 cycled in manual mode and positioned as required.

10. Nuclear Instrumentation check completed per SOP.

l l a. The following trip values were obtained during the check:

IRM-2, run-i n seconds (11+1) Scram seconds (9+1)

IRM-3, run-i n seconds (1171) Scram seconds (9T1)

WRM-4, run-in  % (114+1) - Scram  % (119+1) ~~

PRM-5, run-in  % (11471) Scram  % (11971)

PRM-6, run-in  % (11471) Scram  % (11971)

11. Channel 4, 5, and 6 pots returned to last heat balance position. ""

12. SRM-1 detector response checked and set to indicate > 1 cps.

Rev. 12/16/86 App'd MA SOP /A-la

l TOR STARTUP CHECKSHEET, FULL POWER OPERATION (Cont'd)

13. Check of process radiation monitors (f ront panel checks).

a. Fission prcduct monitor.

b. Secondary coolant monitor.

NOTE: Items 14 through 35 are to be completed in sequence immediately ,

prior to pulling rods for a reactor startup. j

] 14.. Annunciator tested.

15. Annunciator alarm cleared or noted.

16. Power selector switch ISB in position required.  :

[

~

'. 7 . a. Bypass switches 2S40 and 2541 in position required. {"

b. All keys removed f rom bypass switches.

18. Master switch 1S1 in "on" position.

19. Magnet current switch on, check " Reactor On" lights.

" 20. Reactor isolation, f acility evacuation and ARMS checks (af ter maintenance day).

These items are to De checked with scrums and rod run-ins reset, and when appropriate items are actuated, verify that the TAA's do trip.

] a. Reactor isolation switch (leave valves and doors closed) (after maint. day),

b. Facility evacuation switch (check outer containment horns) (after maint. day). ,

c. ARMS trip setpoints checked and tripped, check buzzer operational locally i for all channels and remotely for channels 1 through 4 and 9.

Channel 1 - Beam Room South Wall i Channel 2 - Beam Room West Wall Channel 3 - Beam Room North Wall "I

Channel 4 - Fuel Storage Vault Channel 6 - Cooling Equipment Room 114 Channel 7 - Building Exhaust Air Plenum (after maintenance day). l Channel 8 - Reactor Bridge (switch in " Normal") (after maintenance day).

9 Channel 9 - Reactor Bridge backup (switch in " upscale") (after maint. day).

d. Check HV readings: volts. (520 + 10 VDC)

e. Check 150V reading: volts. (150 VDC +20)

_ ( - 5)

f. Selector switch on ARMS in position 5.

g. Trip backup monitor with attached source (after maintenance day).

h. Reactor isolation horns switch in " Isolation Horns On" position. Valves and j doors open. j

] i. All ARMS trips set per S0P. l

j. Check ventilation fans, containment and backup doors. I

21. Operate reg blade from full-out to full-in and set at 10" + .05".

a. Check rod run-in function at 10% . withdrawn and annunciator at rod bottomed.

22. Raise blade A to 2" and manually scram.

23. Raise blade B to 2" and trip manual rod run-in. ,

[ 24 Raise blade C to 2" and scram by WRM trip. l

25. Raise blade D to 2" and scram by IRM trip. i 26 Annunciator board energized; horn on.

[ 28.27. Jumper and tag log cleared or updated.

IRM recorder in fast speed.

[ 29.30 Check magnet current for 90 ma on each magnet.

Cycle WRM range switch.

)

[ 31. Predicted critical blade position ( inches).

[ 32. Pre-startup process data taken.

E 33. Visually check room 114 and D.I. area after all systems are in operation.

[ a. Check oil reservoir for pump P501A, P501B, and P533 for adequate supply.

Add if necessary. '

[34 Routine patrol completed.  ;

,35. All reactor and license related systems upon which maintenance was performed have ]j been reviewed and are operable. ]l 1 36. Reactor ready for startup.

Time (Completed)

Shif t Supervisor / Acting Shift Supervisor ]

'. 12/16/86 App'd th- SOP /A-lb

're Monday Startup Pot Settings D Last Sheet 4 5 6

)

D This Sheet 1 Pot Changes Sal MWD Date Time ChanneT Setting 01 Element Record

>re Position Fuel Element MWD to Date;

)

MO _ ]

MO- ]

1 M0-M0-

]

)

i M0- )

i MO ,. ]

MO- - ]

1 M0- __ ]

cer Date " Time Time Time Time at This g Arrival Departure Hrs Min Pwr Level (days) MWD Comments

^

1 i

l I

I i

l l

l s l l sv.12/16/86 App'd M 50P/A-6b I

REVISION NUMBER 5 Site Emergency Procedures Section (SEP) ,

to EMERGENCY PROCEDUKES Issued January 8, 1985 v

Sec. No. Page No. Date Revised SEP-1 1 2/10/87 SEP-1 3 2/10/87 SEP-1 4 2/10/87 SEP-2 1 2/10/87 SEP-5 2 2/10/87 SEP-6 2 2/10/87 SEP-7 2 2/10/87 SEP-8 1 2/10/87 ,

SEP-9 1 2/10/87 SEP-11 1 2/10/87 SEP-11 2 2/10/87 SEP-11 3 2/10/87 SEP-11 4 2/10/87 EMERGENCY 2/10/87 CALL LIST 9

l 11-4

Page 1 of 5 SEP-1  ;

ACTIVATION OF FACILITY EMERGENCY ORGANIZATION PROCEDURE 1

1. FACILITY EMERGENCIES: )

j

1. The Facility Emergency Organization (FE0) shall be activated as per  !

the Facility Emergency Procedures (FEP) for each FACILITY ' EMERGENCY f CLASSIFICATION: l a) Facility Evacuation b) Reactor Isolation ,

c) Fi re j d) Medical e) Security [ SAFEGUARDS INFORMATION AS PER 10CFR50.34(c)]  ;

2. In addition to the procedures for each FACILITY EMERGENCY CLASSIFICATION, an assessment of offsite consequences shall be determined.

II. EMERGENCIES WITH POSSIBLE OFFSITE CONSEQUENCES:

1. The duty shift supervisor or individual authorized to assume the Emergency Director position shall determine the need to activate the FE0 for emergencies with possible of fsite consequences. This need will  ;

be based upon the action levels specified for the mini...Jm Classifica- ,

tion, UNUSUAL EVENT. These action levels are: 3

a. Report or observation of severe natural phenomenon.

b. Threats to or breaches of security. 3  ;

c. Concentration of airborne radioactivity at the stack monitor ,

exceeding 3800 MPC when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.  !

d. The projected concentration of airborne radiological effluents at the distance corresponding to the nearest site boundary 2xceeding 10 MPC when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e. Prolonged fire or explosion within the facility that can result in a release of radioactivity that would cause exposures of the public or staff approaching 1 rem whole body or 5 rem thyroid.

l

f. Other plant conditions exist that warrant assuring emergency person- l nel are available to respond to an emergency to prevent exposures of {

1 rem whole body or 5 rem thyroid to the public or staff. I J

Rev. 2/10/87 App'd idW\

4

SEP-1 (Cont'd) Page 3 of.5

8. The EMERGENCY DIRECTOR shall:

i a. Evaluate the information provided by the individual activating the FE0 to initially classify the emergency as per TABLE I, EMERGENCY-CLASSES.

Time Basis Classification

b. Direct the actions detailed in the procedure for the appropriate emergency classification.

UNUSUAL EVENT SEP-2 ALERT SEP-3 SITE AREA EMERGENCY SEP-4 NOTE: The EMERGENCY DIRECTOR must continually ]

evaluate the information received throughout ]

an emergency to determine if an escalation ]

of emergency class is warranted. ]

Rev. 2/10/87 App'd NM

ISEP-l'(Cont'd)~ Page 4 of 5 TABLE I ,

ACTION LEVELS FOR EMERGENCY CLASSES Unusual Event

1. P.eport or observation of severe natural phenomenon.

2. Threats to or breaches of security. ]

3. Concentration of airborne radioactivity at the stack monitor exceeding 3800 MPC when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4. The projected concentration of airborne radiological effluents at the distance corresponding to the nearest site boundary exceeding 10 MPC when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

5. Prolcnged fire or explosion within the facility that can result in a release of radioactivity that would cause exposures of the public or staff approaching 1 rem wbcle body or.5 rein thyroid.

6. Other plant conditions exist that warrant assuring emergency personnel are available to respond to an emergency to prevent exposures of 1 rem whole body or 5 rem thyroid to the public or staff.

Alert

1. Concentration of airborne radioactivity at the stack monitor exceeding 19,000 MPC when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

2. The projected concentration of airborne radiological effluents at the distance corresponding to the nearest site boundary exceeding 50 MPC when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3. Radiation levels at the distance corresponding to the nearest site boundary of 20 mrem /hr for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> whole body or 100 mrem thyroid dose.

4. Loss of physical control of the facility.

5. Other plant conditions exist with a level of significance of a major failure of fuel cladding but primary and containment boundaries exist to reduce releases.

Rev. 2/10/87 App'd M

Page 1 of 5 ,

SEP-2 UNUSUAL EVENT PROCEDURE ACTIONS LEVELS:

E there is:

a. Report or observation of severe natural phenomenon.

b. Threats to' or breaches of ' security. ]

(See REACTOR EMERGENCY PROCEDURE (REP-22)

c. Concentration of airborne radioactivity at the. stack monitor exceed- {

ing 3800 MPC when' averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. <

i NOTE: USE OVERLAY I TO DETERMINE EXTENT OF ACTIVITY FOR I0 DINE AND PARTICULATE. USE OVERLAY II TO DETERMINE EXTENT OF GASE0US ACTIVITY.

d. The projected concentration of airborne radiological effluents at the distance corresponding to the nearest site boundary exceeding 10 MPC when averaged over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

e. Prolonged fire or explosion within the facility that can result in a release of radioactivity that would cause exposures of the public or staff approaching 1 rem whole body or 5 rem thyroid.

f. Other plant conditions exist that warrant assuring emergency personnel are available to respond to an emergency to prevent exposures of 1 rem whole body or 5 rem thyroid to the public or staff.

THEN at least an UNUSUAL EVENT condition exists.

IMMEDIATE ACTIONS:

1. Activate the Facility Emergency Organization (FEO), as per ACTIVATION OF FACILITY EMERGENCY ORGANIZATION PROCEDURE if not already activated.

2. Operations shall provide information to the EMERGENCY DIRECTOR / EMERGENCY COORDINATOR.

3. If airborne activity is involved, contint.e with step 4. E not, go to step 7.

4. Time and date stack monitor charts for reference.

Rev. 2/10/87 App'd \dhM_

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_ _ _ _ _ _ _ _ _ _ . _ J

Page 1 of 3 SEP-8 EMERGENCY EQUIPMENT MAINTENANCE PROCEDURE

Purpose:

This procedure states how the operational readiness of emergency equipment and supplies required by the Emrgency Procedures is to be maintained, calibrated, tested, and periodically inventoried.

Health Physics' Emergency Equipment:

Health Physics' Emergency Equipment and supplies required by the Site Emergency Procedures are maintained in two designated readily accessible locations.

The two locations are the MURR Emergency Equipment Cabinets in the copy machine room and the Backup Emergency Equipment Cabinet at RPDB in room 10. The inventory check lists for each location are attached to this procedure and are used to verify the contents of the cabinet quarterly. All maintenance and inventory records on ]

emergency equipment shall be maintained in the Health Physics Office. ]

The civil defense kits stored at each location contain a G-M survey meter, an ion chamber meter (500 R/hr range), and a set of self-reading pocket dosimeters.

These kits are exhanged annually by Missouri SEMA for updated kits of instruments which have been calibrated by SEMA. The self-contained breathing apparatus sets are inspected by the Columbia Fire Department annually. The emergency Health Physics' equipment is supplemented by the Health Physics' instruments and equipment used for daily routine evaluations and is maintained and calibrated by the Operational Health Physics' Program.

Off-Gas Monitoring Equipment:

The off-gas stack monitor has thcae channels: gas, particulate, and iodice.

The three channels are calibrated semiannually by the Reactor Chemistry group using Reactor Chemistry Standard Operar;ing Procedures:

RC/III-2 Calibration of Stack Particulate Monitor RC/III-3 Calibration of Stack Iodine Monitor RC/III-4 Calibration of Stack Gas Monitor The operability of the off-gas monitor is checked as part of the start-up check-list for normal reactor start-ups.

Fire Fighting Equipment:

The fire hydrant in the reactor site is checked operable annually by the l Columbia Fire Department. The fire ext'nguishers are checked semiannually by j the University of Missouri Physical Plart.

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Rev. 2/10/87 App'd M

Page 1 of 2 SEP-9 TRAINING PROCEDURE FOR EMERGENCY PREPAREDNESS

1. TRAINING OF EMERGENCY ORGANIZATIONS A. FACILITY EMERGENCY ORGANIZATION (FE0) TRAINING This organization consists of MURR staff in the Director's Office, ]

Operations, Health Physics and Reactor Chemistry groups. This organi-zation will respond to both GENERAL FACILITY EMERGENICES (FACILITY EVACUATION, REACTOR IS0UTTON, FIRE, MEDICAL AND SECUPITY EMERGENCIES) and EMERGENCIES WITH POSSIBLE OFFSITE CONSEQUENCES (UNUSUAL EVENT, ALERT, ]

SITE AREA EMERGENCY, PARTIAL SITE AREA EVACUATION).

1. The members of the FE0 will train initially and annually thereafter, by emergency plan and procedure review of each member's role in emergency preparedness. This training will be documented by each member signing the EMERGENCY PLAN / PROCEDURES REVIEW DOCUMENTATION ]

LIST.

2. Annual onsite emergency drills shall be conducted as action drills to test the training of FE0 members to carry out their roles under simulated emergency conditions.

B. TRAINING 0F FACILITY STAFF OTHER THAN FE0 MEMBERS j

1. The members of MURR staff not assigned to the FE0 shall initially and annually thereafter be trained as to their respective actions for each site and facility emergency classification. This training may be by seminar, lecture, or video tape sessions.

2. Annual onsite emergency drills will tes these members' ability to properly respond to simulated energency conditions.

C. TRAINING OF EMERGENCY SUPPORT ORGANIZATIONS

1. The members of EMERGENCY SUPPORT ORGANIZATIONS shall be trained initially and biennially thereafter on their role in maintaining Emergency Preparedness. This will be performed by discussions between MURR staff and the members of each SUPPORT ORGANIZATION, stressing familiarization with the facility or changes to the emergency plan or procedures.

2. Biennial emergency drills shall be conducted to test, as a minimum, the communication link and notification procedures with these EMERGENCY SUPPORT ORGANIZATIONS.

Rev. ?/10/87 App'd b

Page 1 of 4 ]

SEP-11 MONITORING PLANNED EXPOSURES IN EXCESS OF LIMITS IN 10 CFR 20 NOTE: THE EMERGENCY DIRECTOR MUST APPROVE EACH VOLUNTEER TO EXCEED EXPOSURE LIMITS OF 10 CFR 20.

PURPOSE:

This procedure provides the guidelines for monitoring, dosimetry and F records for persons who volunteer for life-saving and accident mitigating activities which could result in planned radiation exposures in excess of 10 CFR 20 limits.

GUIDELINES:

A. Emergency Exposure Guidelines

1. A volunteer may be authorized to receive up to 100 Rem to save human life.

2. A volunteer may be authorized to receive up to 25 Rem to save vital reactor equipment and/or to prevent radiation exposures t,o merabers of the general public in excess of the following Protect. ion Action Guides (PAGs).

The Protective Actions for all classifications are based upon ::

1 l PAG of 1 rem dose equivalent for whole body and 5 rem dose equivalent tnyroid to members of the general public and MURR staff on site.

B. Personnel Selection Considerations

1. Individual is a volunteer (preferably 45 years or older).

2. Individual is familiar with the radiological consequences of emergency radiation exposures.

3. Women of childbearing age shall not take part (Reg. Guide 8.13).

Rev. 2/10/87 App'd hh ,

l SEP-11 (Cont'd) Page 2 of 4 )

C. Self-reader Dosimeters and Locations Type Range mr Location PC 0 - 200 Receptionist Desk - Lobby PC 0 - 200 Health Physics Office PC 0 - 200 Film Badge Racks PC 0 - 200 Emergency Lockers PC 0 - 200,000 Health Physics Office, (Instrument Cabinet in Health Physics Closet) 415A 0 - 9999 Health Physics Office, (lastrument Cabinet in Health Physics Closet)

NOTE: PC = Self-reader Pocket Chamber 415A = XETEX-415A Digital Dosimeter PROCEDURE:

NOTE: Although it is preferable to perform and document the steps presented below prior to emergency workers receiving the exposure, the situa-tfon (e.g., life saving measures) may require the EMERGENCY DIRECTOR to verbally authorize the exposure and ensure completion of docu- I mentation when time permits.

1. Ensure that the dose rate in the task area is known or measurable.

2. Ensure that volunteer has assigned film badge.

3. Complete Worksheet E giving careful attention to item C considerations. ]

4. Assign the volunteer 6 self-reader pocket chamber or equivalent immediate readout monitor with a range double the anticipated total exposure for the assignment.

5. The Emergency Director or a delegate shall obtain the exposure recorded by the readout device at a frequency that assures adequate control to prevent exposures in excess of the planned limits for the volunteer,

a. The exposure received at each reading shall be recorded on Worksheet E.

FOLLOW-UP: ]

1. Expedite processing of volunteer's film badge to verify total exposure as indicated by self-reading dosimeters. 1 i

Rev. 2/10/87 App'd ([Mfy\ '

t

SEP-11 (Cont'd) Page 3 of 4 ]

2. If individual's dose equivalent exceeds any 10 CFR 20 limits, they should be examined by a physician to determine the need, extent and nature of medical surveillance or treatment.

REPORTING REQUIREMENTS:

Complete reporting requirements of 10 CFR 20.403.

1 Rev. 2/10/87 App'd ($MA

SEP-11 (Cont'd) Page 4 of 4 )

WORKSHEET E ,

l RECORD AND AUTrl0RIZATION OF PLANNED EXPOSURES j EXCEEDING LIMITS IN 10 CFR 20 q A. - Emergency Exposure Authorization:

EMERGENCY DIRECTOR DATE i B. Volunteer Signature: I

- DATE C. Recerd of Exposure Measured by Dosimeters

1. Name l

2. Exposure limit assigned for the job mrem.

24 Hour Dos. Dos. Unused Comment 24 Hour Dos. Dos. Unused Comment Time Type mr mr Time Type mr mr i

If 4

Rev. 2/10/87 App'd \Mh4k

SECTION III REVISIONS TO THE HAZARDS

SUMMARY

REPORT 1 July 1986 through 30 June 1987 HAZARDS

SUMMARY

REPORT (original July 1,1965)

There were no revisions to the Hazards Summary between 1 July 1986 and 30 June 1987.

1 III-1 q 1

SECTION IV PLANT AND SYSTEM MODIFICATIONS 1 July 1986 through 30 June 1987 NOVEMBER 1986 Modification 86-4: This modification replaced the original 30' graphite wedge installed in position 4 with an element of new design that can be utilized as an irradiation facility.

The new graphite wedge was constructed to the original wedge specifications except for the addition of two sample irradiation p3sitions, one two inch diameter and one three inch diameter. The new wedge employs nuclear grade graphite which exceeds the specifications of the original wedge graphite.

The new graphite wedge is designed to allow its removal without backing the beamport "C" beam tube through the biological shield. The removal of the original 30' wedge required the beamport tube to be backed 10" through the biological shield.

Measurements of the reactivity effect of inserting and removing sample holders, as well as the reactivity effect of the wedge change were made subsequent to the replacement. The new element was installed November 6,1986.

APRIL 1987 Modification 86-3: This modification is designed to replace the original Pneumatic Tube system control circuitry with circuitry utilizing more current electrcq1c devices. This change was brought about by the incr eased frequency of system component failures.

This modification eliminates much of the previous relay and switch logic, and should make the system more reliable. It involves removing the existing con-trol boxes at each dispatch / return station and installing new boxes with circuitry designed specifically for current and future needs.

IV-1

SECTION V NEW TESTS AND EXPERIMENTS 1 July 1986 through 30 June 1987 New experimental programs during this period are as follows:

RUR-181 EXPERIMENTER: M. Carter /S Gunn DESCRIPTION: An addendum was added to the RUR authorizing the production of compact sealed sources of Gadolinium-153 for medical research dealing with the diagnosis of osteoporosis in humans.

RUR-234 EXPERIMENTER: 0. M. Alger DESCRIPTION: An addendum was added to the RUR authorizing the irradiation of targets to be used in the production l

of sealed sources to be used in portable radiographic and bone densitometry equipment.

RUR-256 EXPERIMENTER: G. Schupp/B. Yelon DESCRIPTION: An Addendum was added to the RUR authorizing the irradiation of new sealed sources for high intensity kossbauer scattering experiments.

V-1

SECTION VI SPECIAL NUCLEAR MATERIAL ACTIVITIES 1 July 1986 through 30 June 1987

1. SNM Receipts: A total of 19 new fuel elements were received from Babcock and Wilcox (B & W), Lynchburg, Virginia.

Grams Grams Shipper Elements U U-235 B&W M0226, M0227, M0228, M0229, M0230, 15,738 14,660 M0231, M0232, M0233, M0234, M0235, M0236, M0237, M0238, M0239, M0240, M0241, M0242, M0243, M0244, M0246, M0247, M0248

2. SNM Shipments: A total of 24 spent fuel elements were shipped to Savannah River Plant, Aiken, South Carolina, for reprocessing.

Grams Grams Shipper Elements U U-235 MURR M058, M0103, M0105, M0106, M0107, M0108, 16,118 14,065 M0109, M0111, M0112, M0113, M0114, M0115, M0116, M0117, M0122, M0125, M0127, M0132, M0134, M0136, M0137, M0138, M0140, M0144

3. Inspections: There were no routine safeguards inspections conducted by the Nuclear Regulatory Commission (NRC), Region III office during the time period 1 July 1986 to 30 June 1987. The MURR Special Nuclear Material Control Pro-cedures were reviewed by the Procedures Review Subcommittee (of the Reactor Ady'isory Committee) as per the annual requirement.

4. SNM Inventory: As of 30 June 1987, MURR was financially responsible for the following DOE owned amounts:

Total V = 42,807 grams Total U-235 = 38,298 grams Included in these totals are 36 grams of U and 34 grams of U-235 non-fuel, 00E owned. In addition to these totals, MURR owns 156 grams of U and 74 grams of U-235. All of this material is physically located at the MURR.

VI-I

Fuel elements on hand have accumulated the following burnup as of 30 June 1987:

Fuel Element Accumula ted Fuel Element Accumulated Fuel Element Accumula ted Number Megawatt Days Number Megawatt Days Number Megawatt Days .,

M0133 148.47 M0210 137.33 M0230 74.22 M0135 148.47 M0211 112.58 M0231 75.47 M0141 146.83 M0212 149.98 M0232 64.52 M0142 145.25 M0213 112.58 M0233 75.47 M0143 99.72 M0214 149.98 M0234 64.52

-~ ~~~

M0145 148.27 M021s 89.28 M0235 64.92 M0146 148.66 M0216 103.68 M0236 51.58 M0147 148.27 M0217 86.39 M0237 64.92 M0148 148.66 M0218 103.68 M0238 51.5 M0149 146.83 M0219 146.56 M0239 58.39 M0200 149.82 M0220 110.96 M0240 58.39 M0201 147.56 M0221 146.56 M0241 0 ,

M0202 149.82 M0222 110.96 M0242 17.98 M3203 147.56 M0223 89.28 M0243 7.45 l

l M0204 148.17 M0224 98.01 M0244 17.98 l

M0205 86.39 M0225 98.01 M0245 7.45 M0206 154.76 M0226 99.72 M0246 0 M0207 149.28 M0227 80.91 M0247 0 M0208 144.58 M0228 74.22 M0248 0 M0209 149.28 M0229 80.91 Average Burnup = 98,19 MWD ,

VI-2

SECTION VII REACTOR PHYSICS ACTIVITIES 1 July 1986 through 30 June 1987

1. Fuel Utilization: During this period, the following elements reached their licensed burnup and were retired.

Serial Number Final Core Date Last Used MWDs M0133 86-47 9-11-86 148.47 M0135 86-47 9-11-86 148.47 M0141 87-16 4-16-87 146.83 M0145 86-65 12-11-86 148.27 P0146 86-68 12-24-86 148.66 M0147 86-65 12-11-86 148.27 M0148 86-68 12-24-86 148.66 M0149 87-16 4-16-87 146.83 M0200 86-59 11-05-86 149.82 M0201 86-39 7-24-86 147.56 M0202 86-59 11-05-86 149.82 M0203 86-39 7-24-86 147.56 M0204 87-21 5-06-87 148.17 M0206 87-24 5-21-87 149.36 M0207 87-7 2-19-87 149.28 M0208 87-24 5-21-87 144.58 M0209 87-7 2-19-87 149.28 M0210 87-22 5-08-87 137.33 M0212 87-20 4-30-87 149.98 M0214 87-20 4-30-87 149.98 M0219 87-25 5-28-87 146.56 M0221 87-25 5-28-87 146.56 VII-1

Due to the requirement of having less than 5 kg of unirradiated fuel in possession, initial criticalities are obtained with four new elements or fewer as conditions dictate. A core designation consists of eight fuel elements of which only the initial critical fuel element serial numbers are listed in the following table. To increase operating efficiency, fuel ele-ments are used in mixed core loadings. Therefore, a fuel element fabrication core number is different from its core load number.

Fabrication Serial Initial Core Initial Core No. No. Load Number Operating Date 37 M0143 86-44 8-21-86 44 M0226 86-44 8-21-86 44 M0227 86-55 10-09-86 44 M0228 86-56 10-16-86 44 M0229 36-55 10-09-86 44 M0230 86-56 10-16-86 44 M0231 86-68 12-20-86 45 M0232 87-1 1-08-87 45 M0233 86-68 12-20-86 45 M0234 87-1 1-08-87 45 M0235 87-7 2-12-87 45 M0236 87-10 3-05-87 45 M0237 87-7 2-12-87 45 M0238 87-10 3-05-87 45 M0239 87-13 3-26-87 46 M0240 87-13 3-26-87 46 M0242 87-21 4-30-87 46 M0244 87-21 4-30-87 VII-2

2. Fuel Shipping: Twenty-four spent fuel elements were shipped from MURR to Savannah River Plant, Aiken, South Carolina. The identification numbers of these elements are:

M058 M0109 M0116 M0134 M0103 M0111 M0117 M0136 M0105 M0112 M0122 M0137 M0106 M0113 M0125 M0138 M0107 M0114 M0127 M0140 M0108 M0115 M0132 M0144

3. Fuel Procurement: Babcock and Wilcox, Lynchburg, Virginia is MURR's current fuel assembly fabricator. This work is contracted with the U. S. Department of Energy and administered by E G & G Idaho, Idaho Falls, Idaho. As of 30 June 1987, forty nine fuel assemblies fabricated by B & W had been

( received and forty five used in cores at 10 MW.

4. Licensing Activities: A request for an increase in Special Nuclear Material Inventory under our Facility License submitted in December, 1982 is pending.

A revision to Technical Specifications 4.4.d requiring two operating paral-lel pool pumps submitted in February,1982 is pending. A request for an increase in the possession limit for depleted uranium, dated May,1986, was withdrawn. A revision to Technical Specifications which involves changes l which are editorial in nature was submitted November 22, 1985 and is pend-ing. A submittal was made September 12, 1986 pertaining to the new MURR fuel design and associated revisions to the Technical Specifications. The fuel design is currently being evaluated by Nuclear Regulatory Commission's subcontractor. Therefore, this submittal is also pending.

VII-3

5. Reactor Characteristic Measurements: Sixty six refueling evolutions were comple ted. An excess reactivity verification was performed for each refuel-ing and the average excess reactivity was 2.6%. MURR Technical Specifica-tion 3.1(f) requires that the excess reactivity be less than 9.8%.

Reactivity measurements were performed for 16 evolutions to verify reactivity parameters for the flux trap. Six shim and regulating blade calibrations were performed. Six reactivity measurements were made on graphite reflector sample movements.

A physical inspection of the following fuel elements was performed to verify the operational parameters.

M0200 from Core 41 during October 1986 M0208 from Core 42 during March 1987 All measurements were within operational requirements.

VII-4

SECTION VIII

SUMMARY

OF RADI0 ACTIVE EFFLUENTS RELEASED TO THE ENVIRONMENT Sanitary Sewer Ef fluent 1 July 1986 through 30 June 1987 Descending Order of Activity Released Nuclide Amount (Ci! Nuclide Amount (Ci)

H-3 1.11E-01 Tm-170 7.99E-05 S-35 9.97E-03 Eu-152 7.61E-05 Sb-124 1.56E-03 Sb-122 4.09E-05 Cr-51 4.97E-04 Rh-105 3.55E-05 Te-125M 4.63E-04 Re-188 2.82E-05 Co-60 3.35E-04 Cs-137 2.36E-05 Cu-64 3.28E-04 Rb-86 2.18E-05 Zn-65 2.83E-04 Sb-125 1.72E-05 Re-186 2.58E-04 Fe-59 1.52E-05 As-77 1.84E-04 Gd-159 1.35E-05 Ru-105 1.62E-04 Ba-131 1.34E-05 Ag-110M 8.68E-05 Na-24 1.27E-05 Sc-46 8.64E-05 Eu-154 1.17E-05 VIII-1

Stack Effluent kJuly1986through30 June 1987 Descending Order of Activity Released for Isotope Levels Greater than 10-4 Ci Nuclide Amount (Ci)

Ar-41 7.43E+02 H-3 1.03E+01 I-133 8.00E-04 C1-38 7.47E-04 I-135 6.72E-04 I-131 5.64E-04 As-77 4.88E-04 ,

Hg-197 2.96E-04 I-134 2.92E-04 Xe-135M 2.31E-04 I-128 2.10E-04 I-132 1.97E-04 Sr-89 1.97E-04 Stack flow rate = 16,200 ft / 3min. ' '

f' ,

VIII-2

SECTION IX

SUMMARY

OF ENVIRONMENTAL SURVEYS ,

1 July 1986 through 30 June 1987 Environmental samples are collected two times per year at eight locations and analyzed for radioactivity. These locations are shown in Figure 1. Soil and vegetation samples are taken at each location. Water samples are taken at three of the eight locations. Results of the samples are shown in the following tables.

1. Sampled during October 1986.

Detection Limits Matrix Alpha Beta __ Gamma Tri tium Water 0.73 pCi/l 1.72 p;i/1 617 pC1/1 12.0 pCi/ml Soil

• 0.43 pCi/g 1.01 pCi/g 5.90 pCi/g ----

Vegetation

• 1.24 pCi/g 2.89 pCi/g 9.25 pCi/g 2.46 pCi/g

• Gamma and tritium analyses are basti on wet weights, a & S analyses are based on dried weights.

Determined Radioactivity Levels Vegetation Samples Satple Alpha Beta Gamma Tri tium (pCi/g) (pC1/g) (pCi/g) (pCi/g) 1-V-30 < 1.24 9.06 34.56 < 2.5 2-V-30 < 1.24 15.19 26.86 < 2.5

- 3-V-30 < 1.24 34.38 < 9.26 < 2.5 4-V-30 < 1.24 22.51 < 9.26 < 2.5 5-V-30 < 1.24 9.77 < 9.26 < 2.5 6-V-30 < 1.24 9.10 < 9.26 < 2.5 7-V-30 < 1.24 12.44 < 9.26 < 2.5 10-V-30 < 1.24 31.81 < 9.26 <- 2.5 IX-1

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Determined Radioactivity Levels Soil Samples Sample Alpha Beta Gamma (pCi/g) (pCi/g) (pCi/g) 1-S-30 < 0.43 9.02 17.61 2-S-30 0.58 6.43 32.04 3-S-30 < 0.43 7.52 40.84 4-S-30 0.58 5.91 35.09 5-S-30 < 0.43 9.44 36.60 6-S-30 < 0.43 2.77 15.49 7-S-30 < 0.43 6.75 19.76 10-S-30 0.88 9.50 18.50 Determined Radioactivity Levels Water Samples Sample Alpha Beta Gamma

• Tri tium (pCi/1) (pCi/1) (pCi/1) (pCi/ml) 4-W-30 < 0.73 4.09 < 617 < 12.0 6-W-30 < 0.73 2.48 < 617 < 12 0 10-W-30 < 0.73 9.37 1864** < 12.0

• Second water sample set was druwn November 1986 for analysis due to back-ground fluctuations for original water sample gamma analysis resulting in high detection limit.

• • Gross activity necessitated isotope identification for sample 10W30 using Reactor Chemistry Ge(L1) detector. Sample containea I-131 = 341 pCi/1 and Tc-99m ='91.7 pCi/1, both of which came from the Medical Center at the University of Missouri-Columbia. The University of Missouri Research Reactor does not process these isotopes.

l l

IX-3

2. Sampled during April 1987.

Detection Limits Matrix Alpha Beta Gamma Tri tium Water 0.63 pCi/1 1.57 pC1/1 206.4pCi/1 12.4 pCi/ml Soi1* 0.46 pC1/g 1.15 pCi/g 1.92 pC1/g N/A Vegetation

• 1.56 pCi/g 3.87 pCi/g 3.44 pCi/g 10.20 pCi/g

• Gamma and tritium analyses are based on wet weights, a and C analyses are based on dried weights.

Determined Radioactivity Levels Vegetation Samples Sample Alpha Beta Gamma Tri tium (pCi/g) (pCi/g) (pCi/g) (pCi/g) 1-V-31 < 1.56 15.85 3.53 < 10.20 2-V-31 < 1.56 21.81 < 3.44 < 10.20 3-V-31 < 1.56 20.34 < 3.44 < 10.20 4-V-31 < 1.56 29.95 < 3.44 < 10.20 5-V-31 < 1.56 34.78 < 3.44 < 10.20 6-V-31 < 1.28* 34.53 < 3.44 < 10.20 7-V-31 < 1.56 22.55 < 3.44 < 10.20 10-V-31 < 1.56 23.03 < 3.44 < 10.20

• Re-analysis of sample resulted in lower detection limit.

IX-4

Determined Radioactivity Levels Soil Samples Sa.mpl e Alpha Beta Gamma (pCi/g) (pCi/g) (pCi/g) 1-S-31 < 0.46 11.67 7.77 2-S-31 0.95 12.48 9.21 3-S-31 0.49 9.85 7.54 4-S-31 0.68 8.86 5.54 5-5-31 0.87 11.74 5.82 6-S-31 < 0.46 7.89 4.87 7-5-31 0.77 11.58 8.36 10-S-31 0.98 12.06 8.49 Determined Radioactivity Levels Water Samples Sampic Alpha Beta Gamma Tri tium (pC1/g) (pCi/g) (pCi/g) (pCi/g) i 4-W-31 < 0.63 4.92 < 206.4 < 12.4 j 6-W-31 < 0.63 5.95 < 206.4 < 12.4 10-W-31 < 0.63 11.67 < 206.4 < 12.4 Environmental samples were also collected around Sinclair RAD WASTE Facility at four locations. Soil and vegetation samples were taken at each location.

Results of these samples are shown in the following tables.

1. Sampled during October 1986.

Detection Limits Matrix Alpha Beta Gamma Tri tium j Soi1* 0.56 pCi/g 1.39 pCi/g 1.83 pC1/g N/A Vegetation

• 1.70 pC1/g 4.22 pC1/g 3.31 pCi/g 9.84 pCi/g

• Gamma and tritium analyses are based on wet weights, a & S analyses are based on dried weights. ,

IX-5 i l

Determined Radioactivity Levels vegetation Samples Sample Alpha Beta Gamma Tritium (pCi/g) (pC1/g) (pCi/g) (pCi/g)

SF-1-V5 < 1.70 13.85 < 3.31 < 9.84 SF-2-V5 < 1.70 22.43 < 3.31 < 9.84 SF-3-V5 < 1.70 26.56 < 3.31 < 9.84 SF-4-V5 < 1.70 17.00 < 3.31 < 9.84 Determined Radioactivity Levels Soil Samples Sample Alpha Beta Gamma (pC1/g) (pCi/g) (pCi/g)

SF-1-SS 0.76 13.90 9.98 SF-2-S5 1.11 13.60 7.65 SF-3-S5 0.67 17.77 9.96 SF-4-SS < 0.56 8.73 4.84

2. Sampled during April 1987.

Detection Limits Matrix Alpha Beta Gamma Tri tium Soi1* 0.36 pCi/g 0.84 pCi/g 5.66 pCi/g N/A Vege tation* 2.05 pCi/g 4.82 pCi/g 8.22 pCi/g 2.2 pCi/g

• Gamma and tritium analyses are based on wet weights, a & B analyses are based on dried weights.

Determined Radioactivity Levels Vegetation Samples Sample Alpha Beta Gamma Tri tium (pCi/g) (pCi/g) (pC1/g) (pC1/g)

SF-1V4 < 2.05 12.96 < 8.22 < 2.2 SF-2V4 < 2.05 16.59 < 8.22 < 2.2 SF-3V4 < 2.05 10.94 < 8.22 < 2.2 SF-4Y4 < 2.05 12.88 < 8.22 < 2.2 IX-6

Determined Radioactivity Levels Soil Samples Sample Alpha Beta Gamma (pCi/g) (pC1/g) (pCi/g)

SF154 0.64 9.74 < 5.66 SFIS4 0.45 9.48 < 5.66 SFIS4 0.45 7.20 < 5.66 SF1S4 0.54 8.04 < 5.66 Radiation and Contamination Surveys The following table gives the number of surveys performed during FY 86-87.

Radiation Surface Contamination Air Samples 1986 July 37 35 23 August 27 24 21 September 66 58 22 October 44 41 24 November 24 16 20 December 48 40 23 1987 January 25 18 22 February 39 34 20 March 32 32 22 April 57 55 22 May 25 22 21 June 26 ,4J 22 450 417 262 TOTALS 58 Radiation Work Permits were issued during the year.

I IX-7

Miscellaneous Items As of April 15, 1987, Orval Olson retired as Manager, Reactor Health Physics, and Susan Langhorst was hired as his replacement. Dr. Langhorst has been a Research Scientist at MURR since 1980, and is also an Assistant Professor of Nuclear Engineering and an ABHP Certified Health Physicist.

Health Physics Procedure #30, EVALUATION OF EXHAUST GAS STACK MONITOR FILTERS IN AN EMERGENCY, was written to provide direction for obtaining samples and data to estimate the emission of radionuclides through the stack exhaust system during an emergency condition at the MURR.

MURR contracted for ADC0 Services, Inc. to act as our institutional waste broker during a two year period beginning in March 1987. Through ADC0, MURR has shipped 7 barrels (total volume = 61.2 cu. f t.) of radwaste on April 30, 1987, and 2 boxes (total volume = 100.8 cu. ft.) of radwaste on June 25, 1987.

A hand exposure reported last year at 23.56 rem in June 1986 was later modified in August 1986 by our personnel dosimetry vendor to be 115 rem in order to reflect exposure to Tm-170 radiation. As a result of this hand dose exceeding 10 CFR 20 limits, MURR received notice of civil pcnalty in January 1987 and paid the penalty in May 1987. To prevent recurrence of the violations cited, changes were made to Health Physics Procedure #16 to better establish activity levels of samples, and Health Physics Procedure #32 was written to define specific health physics procedures for subsequent handling of high Tm-170 activity.

An analysis was performed and operating procedures written for MURR's Can Melter System (CMS) used in reducing volume of radwaste consisting of activated aluminum cans and other aluminum metal wastes. Implementation of the CMS has strengthened MURR's ALARA program in the following ways:

1. For the processing of aluminum irradiation cans, the CMS reduces personnel exposure from 0.91 man-rem / year to 0.14 man-rem / year.

IX-8 .-

2. Processing high radiation level items by imbedding them in the electric can melter crucibles reduces the number of times the radioactive material is handled and eliminates the need to accumulate large volumes of such material for disposal in expensive licensed shipping containers.

3. Because of the significant volume reduction of materials kept in the radwaste storage area close to work areas, total worker exposure is estimated to be reduced by 5 man-rem / year.

The following has been implemented to reduce whole body and hand dose for workers in Radioisotope Applications Group:

1. A variety of Lucite plastic beta shields have been purchased or made to reduce exposures. These include quartz vial openers, laboratory top body (plash) shields, long forceps equipped with plastic hand sheaths, quartz vial loading blocks and caps, quartz vial holders, and syringe shields.

Doses have been reduced by factors ranging from 2 to 10 times original doses.

2. Procedures for decontaminating the outside of irradiated quartz vials were modified to include ultrasonic bath cleaning of the vials to minimize handling of irradiated samples.

3. A new RIA laboratory (Room 244) now houses glove box work involving Re-186 and W-188/Re-188 research. This has spread the amount of glove box work from one laboratory to two, thus reducing congestion, worker exposure from other isotope handling, and possibility of accident and cros s-con tami na tion.

MURR's ALARA efforts have been enhanced by a change in operations' practice (in effect for the past two years) regarding regeneration of pool demineralized resin beds. MURR has found that replacing the depleted beds with new beds rather than regenerating the depleted beds can reduce operations' personnel exposure by IX-9

an estimated 1.2 man-rem / year. An added contribution to ALARA accrues due to a substantial reduction in total activity released to the sanitary sewer, because regeneration solutions were a significant part of the annual sanitary sewer effluent (i.e. Co-60 released for the last two years are a factor of 50 to 100 less than 1984-85).

A member of the Reactor Health Physics staff attended a short course in " Beta Skin Dose Calculations", sponsored by Technical Management Services, Inc., and held in Washington, D. C. on May 18-20. Training included beta point-source dose-rate formulas, recommendation on beta dosimetry, extremity monitoring, evaluating skin contaminations, computer codes for exposure evaluation, recent industry experience, and industry practice in evaluating skin exposures, a

IX-10

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Research Reactor Facility UNIVERSITY OF MISSOURI Research Park Columbia. Missouri 65211 August 17, 1987 Telephone (314) ea2-4211 U.S. Nuclear Regulatory Commission ATTENTION: HerSert N. Berkow, Director Standardization ? Non-Power Reactor Project Directorate Office of Nuclear Reactor ' Regulation Washington, D. C. 20555

REFERENCE:

Docket 50-186 University of Missouri Research Reactor License R-103

SUBJECT:

Annual Report as required by Technical Specification 6.1.h(4).

Dear Sir:

Enclosed are two copies of the Operations Annual Report for the University of Missouri Research Reactor. The reporting period covers 1 July 1986 through 30 June 1987. The remaining twelve copies will be sent in the near future via United Parcel Service.

S ncerely, Walt A. Meyer, r.

Operations Eng neer and Acting Reactor Manager Enclosures (2) xc w/ report: U.S. N.R.C.

c/o Document Control Desk Washington, DC

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lm COLUMBIA iv 4 KANSAS C!TY ROLLA ST. LOUIS <

an BQual Opportunity institution

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