ML20059C505
| ML20059C505 | |
| Person / Time | |
|---|---|
| Site: | University of Missouri-Columbia |
| Issue date: | 06/30/1990 |
| From: | Mckibben J, Meyer W MISSOURI, UNIV. OF, COLUMBIA, MO |
| To: | Weiss S NRC |
| References | |
| NUDOCS 9009050099 | |
| Download: ML20059C505 (172) | |
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EE Research Reactor Facility i
t.NIVERSITY OF MISSOURI !
Research Park {
Columbla. Missouri 45211 Telephone (314) 842 4211 Fax (314 f882 3443 .
August 29.1990 ,
Seymour H. Weiss. Director l l U. S. Nuclear Regulatory Commission
!' PDNP l- M.S. I1-B 20 !
L Washington, D. C. 20555 -
L
REFERENCE:
Docket 50-186 University of Missouri Research Reactor License R-103
SUBJECT:
Anual Repcrt as required by Technical L Specification 6.1.h(4).
Dear Str:
,, Enclosed are two copies of the Reactor Operations Annual Report for the University of Missouri Research Reactor. The reporting period covers 1 July 1989 through 30 Junc .1.990, if you have any questions, please feel free to call.
D Sincerely, ' '
t L
N Walt A. S', eyer, t .
Reactor Manager Enciosure (2)
.xc w/ report: U.S N.R.C.
>' c/o Document Control Desk -
A Washington,'.DC l:
l 000?9 COLUMBIC. XANSAS CITY ROLLA ST. LOUIS 90090$0099 +00630 an equal opponunity institution
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I UNIVERSITY OF MISSOURI -
8 UNIVERSITY OF MISSOURI l RESEARCH REACTOR I >
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- UNIVERSITY OF MISSOURI !
l RESEARCH REACTOR FACILTIY I
REACTOR OPERATIONS l ANNUAL REPORT AUGUST 1990 cl I
'I I Compiled by the Reactor Staff I q Submitted by I Walt A. Meye , r.
Reactor Manager -
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l Reviewed and Appr ved l u
- d. C. McKibben
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g Associate Director II I
I TABLE OF COhrrENTS Section Page Number I. REACTOR OPERATIONS
SUMMARY
I-1 through I-14
, I CHANGES TO THE STANDARD OPERATING PROCEDURES 11-1 through II-6 I III. REVISIONS TO THE HAZARDS
SUMMARY
REPORT III-1 through III-22 IV. PIANT AND SYSTEM MODIFICATIONS IV-1 through IV-6 V. NEW TESTS AND EXPERIMENTS V-1 througli V-2 l .
VI. SPECIAL NUCLEAR MATERIALS ACTIVITIES VI-1 through VI-2 F
g VII. REACTOR PHYSICS ACTIVITIES VII-1 through VII-4 VIII.
SUMMARY
OF RADIOACTIVE EFFLUENTS '
RELEASED TO THE ENVIRONMENT VIII-1 through VIII 3 IX.
SUMMARY
OF ENVIRONMENTAL SURVEYS IX-1 through IX-8 j
,g X
SUMMARY
OF RADIATION EXPOSURES TO i g FACILITY STAFF. EXPERIMENTERS, AND VISITORS X-1 through X-2 l-l I
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SECTION I REACTOR OPERATIONS
SUMMARY
1 July 1989 through 30 June 1990 I The following table and discussion summarize reactor operations in the period 1 July 1989 through 30 June 1990.
I Date Full Power Hour,g Full Power Percent Menawatt Days of Total Time of Schedule
- July 1989 I Aug. 1989 Sept.1989 669.55 673.53 641.25 279.20 280.79 267.34 89.99 90.53 89.06 100.79 101.39 99.75 I Oct. 1989 Nov.1989 609.77 528.61 254.31 220.52 81.85 73.42 91.67 82.23 Dec.1989 681.26 284.08 G1.57 102.56 Jan. 1990 659.64 275.34 88.66 99.30 Feb. 1990 617.80 257.62 91.93 102.97 I Mar.1990 Apr. 1990 May 1990 697.98 634.59 290.91 264.55 93.81 88.26 105.07 98.85 689.04 287.21 92.61 103.73 June 1990 650.13 271.08 90.30 101.13 Total for Year 7753.15 3232.95 88.51% of 99.13% of l Time for Yr.
at 10MW Sched. Time for Yr. at 10 MW l *MURR is scheduled to average at least 150 ho irs oer week at 10MW.
Total time is the number of hours in a month or year.
During the months of October and November (October 29-November 6).
the reactor was shut down for an extended period of time. 7.15 days, to I replace the beryllium reflector. During the maintenance outage, four (4) graphite wedges were also replaced and various in-pool valves were rebuilt.
The extended shutdown accounts for the low percentage of total operating l time during these two months: 81.85% for October and 73.42% for November.
There were 30 unscheduled shutdowns recorded during the year 1 July 1989 through 30 June 1990. Only 8 were in the last six months, which I- 1 I
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could be attributable to the UPS being installed in October 1989. Of these shutdowns,10 were Rod Run Ins (RRis) and 20 were scrams.
Seven of the unscheduled shutdowns (2 RRIs and 5 scrams), were due to a Nuclear Instrument drawer, detector or detector cabling failure generating a I spurious shut down signal (no actual high power or short period was indicated). A bid package is being prepared to upgrade this instrumentation and a request was made to DOE for money to support this upgrade. We l anticipate that integrated cable detectors and new Nuclear Instrumentation will solve the spurious scram problems.
Of the remaining 23 unscheduled shutdowns, nine were component failures, six were due to loss of electrical power to the facility, five were due l to personnel errors, and three were initiated by the duty operator in order to repair equipment.
July 1989 I The reactor operated continuously in July with the following exceptions: five shutdowns for scheduled maintenance and refueling; and six unscheduled shutdowns.
I On July 3, a reactor scram occurred while shifting the wide range I switch upscale during a normal reactor startup. Electronics technicians used a current source to simulate input to the wide range monitor (nuclear instrument channel #4), but could not reproduce this problem. The cause l was suspected to be a sticking relay contact in the feedback network of the picoammeter module for channel #4. The wide range switch was cycled I through its full range and the picoammeter relays operated normally. A normal startup was completed with no further problems, g On July 4, a rod not in contact with magnet rod run-in occurred when control blade "C" disengaged from its magnet during routine shimming. The control blade drive mechanism was removed and the upper guide tube was I realigned to center the amil. The surface of the amil was cleaned. A hot reactor startup was then completed with no further problems. On the next I maintenance day, the offset mechanism and guide tube for control blade "C" was thoroughly inspected and further refinement of the alignment. was accomplished. The control blade was then pull-tested satisfactorily.
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l On July 18, two reactor scrams occurred due to momentary site electrical power interruptions which occurred during a thunderstorm.
I These interruptions were verified by the power plant.
On July 28, the reactor was manually scrammed upon discovery of a ;
- g leak from primary heat exchanger 503A. After noticing an unexplained l decrease in the primary pressurizer water level, the duty shift supervisor
'I immediately made a brief visual inspection of the mechanical eqtJpment room and discovered water leaking from the end cap flange of primary heat exchanger 503A. The control room operator was instructed to immediately scram the reactor, Coincident with the manual scram, the duty operator l
noted that a rod not in contact with magnet RR1 was annunciated and that the natural convection cooling valves (546 A/B) were open. It is believed l that an unannunciated Power Level Interlock scram from the core DP transmitter 929 occurred just shortly before the manual scram duo to a l I pressure dip caused when the leakage from the heat exchanger increast t.
After the manual scram, a Power Level Interlock scram locked in en the I annunciator which was initiated by core DP transmitter 929. I I Subsequent investigation of the heat exchanger revealed a tear in the I end-cap gasket. 'Ihe flange surface was cleaned and the gasket was replaced. The reactor was then returned to normal operation. '
g On July 30, a reactor scram occurred due to a complete loss of site l electrical power. The emergency generator started immediately and assumed its electrical load satisfactorily. The power loss was caused by a l severe thunderstorm. Site power returned after approximately 45 minutes and the reactor was refueled and returned to normal operation.
i E Major maintenance for J 11y included: installing a new bellows seal and bearings on pool demineralizer pump 513B; removing. taking measurernents l on, and reinstalling the Nuclepore irradiator case: and replacing the endcap gasket on primary heat exchanger 503A.
&lgnELlRSS The reactor operated continuously in August with the following exceptions: four shutdowns for scheduled maintenance and refueling; three unscheduled shutdowns: and one power reduction to examine the Beamport I "F" valve arrangement.
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l On August 15, a rod not in contact with magnet rod run-in occurred when control blade "B" disengaged from its magnet during routine 1 I shimming. The drive housing for this control blade was removed, the magnet and anvil vere inspected and the guide tube was realigned.
i i
Electronics technicians subsequently checked and verified the continuity of I l the drive mechanism amphenol connector. A hot reactor startup was completed and the teactor was returned to normal operation.
On August 22, a manual rod run in was initiated to allow investigation of an observed water leak in'the mechanical equipment room. ne small g leak was determined to be coming from a sensing line for primary flow 1 element 912E. The reactor was subsequendy manually scrammed to allow repair of.this line. Machine Shop personnel removed a secuon of tubing I where the leak occurred and placed a compression coupling in the sensing line to rejoin the sections. No more leakage was observed and the reactor was refueled and returned to normal operation. This entire sensing line was I replaced on the next maintenance day.
- l. On August 31, a reactor scram occurred from what was believed to be a spurious reactor loop low flow signal initiated by primary flow element I 912E. The annunciation for this scram was reset before it was adequately documented by the duty operator, therefore the " white rat" scram monitor was used to help determine the most likely cause. It was determined from l scram monitor indications that the most likely cause was a reactor loop low flow signal from primary flow element 912E.
]
One of the sensing lines for this element had been replaced three days previously and it was thought that the transmitter may not have been adequately vented at that time, thus allowing entrapped air to indicate spurious flow oscillations, he transmitter was vented and a i compliance check for that particular scram was completed satisfactorily.
l The reactor was then refueled and returned to normal opam%n.
I Major maintenance items for August included: replacing the 6" discharge line gasket on pool heat exchanger 521B; testing the operability and placing on service the new diesel emergency generator; installing new l bearings and mechanical seal on primary demineralizer pump 513A:
replacing sensing line for primary flow element 912E; installing bismuth /
indium filters in Beamport "F" centertube; removing pool pump 508B to the I Machine Shop for repair; and replacing nuclear instrument channe: #4 detector (wide range monitor) on two separate occasions.
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'I Seotember 1989 1
The reactor operated continuously in September with the following '
Ll exceptions: four shutdowns for scheduled maintenance and refueling: and five unscheduled shutdowns.
On September 3, a reactor scram occurred from what was thought to i be a spurious reactor loop low flow signal initiated by primary flow element ig 912E similar to the unscheduled shutdown of August 31. The annunciation for this spurious signal did not lock in on the annunciator board, so the
" white rat" scram monitor indications were used to help determine the most likely cattse. The transmitter was again vented and a compliance t
check for that particular scram was completed satisfactorily. No further g problems of this nature were experienced.
On September 5, a nuclear instrument channel #4 (wide range l monitor) high power scram occurred during a normal reactor startup. No actual high power condition was indicated on any other instrumentauon.
After troubleshooting efforts, the Electronics technicians subsequently lI replaced the channel #4 compensation potentiometer and the detector and no further problems have occurred.
I On September 0, a reactor scram and isolation was initiated by the ig area radiation monitor exhaust air plenum alarm unit, All personnel exited 3 the containment building as per procedure. All indications from the remote area radiation monitoring station in the Electron!cs Shop read downscale, l ig which indicated a probable electrical problem with the area radiation monitor. A Health Physics technician and the shift supervisor reentered the containment building with a portable radiation monitor and found all l
,l radiation readings in the containment building at normal background. An l Electronics technician determined that a failed resistor in the high voltage '
circuit of the area radiation monitor had caused the scram and isolation.
- The resistor was replaced and the area radiation monitor was tested satisfactorily.
I On September 18, a nuclear instrument channel #5 (power renge l
i I monitor) high power rod run-in occurred shortly after entering automatic control during the completion of a normal reactor startup. Immediately prior to the rod run-in, channel #5 was indicating 109%. 5% below the rod i
g run-in set point of 114% As an operator shimmed the control blades out, channel #5 indication rose to the rod run-in trip set point of 114% before I I-5 I .
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the automatic insertion of the regulating blade compensated. He rod run-in trip point corresponded to an actual power level of 10.4 MW. ne power i level indications on the two other power range monitors were within normal !
5 operating limits. The rod run-in was reset and the indication for channel #5 was lowered to match the other power range monitors before the reactor '
l was returned to normal operation. The operator involved was instructed to be more aware of his power indications while shimming control blades, l On September 25, an anti-siphon system high level rod run-in occurred soon after completion of a normal reactor startup. The rod run-in
.g was a result of a combination of an air leak at a fitting on the anti-siphon i pressure tank and water leaking by the seat of anti-siphon valve 543B. The i valve 543B actuator had been worked on during the maintenance day 1 l' preceding this startup and it appeared that the actuator was not completely closing valve 543B, allowing water to leak by its seat. The combined loss of
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I air pressure and the valve leakage allowed water to accumulate in the a' arm sensing leg which resulted in the rod run in. ne stroke for the valve 543B l was readjusted so that the valve closed properly and the leaking air fitting
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on the anti-siphon tank was tightened. The reactor was then returned to l normal operation.
1 The annual emergency preparedness inspection was conducted by Nuclear Regulatory Commission Inspector, Jim Patterson, from September i 26 through September 28.
j Major maintenance items for September included: replacing inboard / i outboard bearings and mechanical seal on pool pump 508B: replacing compensation potentiometer and detector for nuclear instrument channel i I #4: installing new relay K-58 in the N.I. channel #4 downscale annunciation circult: replacing failed resistor in the high voltage circuit of the area radiation monitoring system; replacing inboard / outboard bearings and )
-l mechanical seal on primary demineralizer pump 513A: rebuilding the actuators on valve 546A (in-pool heat exchanger) and valve 543B (anti-I I siphon).
October 1989 The reactor operated continuously in October with the following exceptions: six shutdowns for scheduled maintenance and refueling; and two unscheduled shutdowns.
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I i On October 10, a nuclear instrument channel #5 (power range l l monitor) high power scram occurred during a normal reactor startup. No j lE W actual high power condition was indicated on any other instrumentation. )
Electronics technicians subsequently discovered, and replaced, a failed dual l trip unit in the drawer for this channel. A normal startup was then '
completed satisfactorily.
l On October 23, a nuclear instrument channel #2 (intermediate range I monitor) period scram occurred during a normal startup when the drywell for this detector was inadvertently bumped while operators were attempting ,
g to physically adjust the drywell for an adjacent detector. The clamps j securing the drywell for this detector were tightened and the operators i involved were instructed to be more careful when working around sensitive !
-l equipment. A normal startup was then completed and the reactor was returned to normal operation.
The reactor was shut down at 2200 on October 29 to commence special maintenance procedure #11 which includes replacement of the !
l rear: tor beryllium reflector. This maintenance outage was successfully l completed and the reactor was returned to full power operadons at 1630 I g November 6,1989.
3 l This extended maintenance outage to replace the beryllium reflector g and four graphite reflectors was accomplished in seven days as opposed to the nine days it took to replace the beryllium reflector in 1981. The l g- experience, tools and knowledge gained during the 1981 beryllium I (3. changeout assisted the planning and preparation for this changeout and resulted in a 9 manrem dose savings (see Annual Report Section IX for l lg Health Physics Summary).
l To change out the beryllium reflector, the reactor pressure vessel had
[l to be dismantled to the split ring flange and to replace the four graphite L reflectors three beam tubes had to be pulled back approximately ten inches.
Lg During the course of the maintenance outage, the valve actuators for i W V507 A/B and V509 were also rebuilt.
l Major maintenance items for the month included: rebuilding the actuators for anti-siphon valves 543A and 543B, and for in-pool heat J
l exchanger valves 546A and 546B: renewing the power connector for control blade "D": replacing the dual trip unit in N.I. channel #5 drawer: removing Beamport "F" centertube: removing the (saddle) air cans in Beamport "A": j 1 '
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l loading spent fuel elements in the National lead fuel shipping cask and transferring them to the beamport floor for temporary storage; replacing I two thrust bearings and one line bearing in pool pump 508A: connecting reactor control instrumentation and back up door power to the new uninterrupuble power supply (UPS) and placing the UPS on service: !
- g rewiring the Elgar AC line conditioner for standby operation and removing from service: and commencing special maintenance procedure #11 which includes replacing the reactor beryllium reflector.
I November 1989 The reactor operated continuously in November following completion l of the extended maintenance outage with the following excepuons: three
.l shutdowns for scheduled maintenance and refueling; and four unscheduled shutdowns.
1 On November 6, a reactor loop low pressure scram occurred during a _
low power (5 kilowatts) test to perform rod worth measurements, when tne l
l primary demineralizer loop isolation valves (527E, 527F) were opened. The demineralizer loop had been isolated and depressurized to change the I demineralizer prefilters. When the isolation valves were opened after completing this task, a slight (momentary) pressure drop was detected at primary pressure transmitter 943 due to its close proximity to the 4
demineralizer inlet isolation valve (527E), thus inidating the scram. This task is usually, though not always, completed with the reactor shut down and ;
the primary system depressurized. In cases where the primary system is l pressurized, the demineralizer prefilters can be isolated locally with manual Isoladon valves without using the air-operated loop isolation valves (527E, I
. I 527F). This results in depressurizing a small portion of the demineralizer loop volume (one filter bank) instead of the entire demineralizer loop. )
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( These local manual valves can then be operated slowly to prevent significant pressure changes. The operators involved in this particular case were instructed to follow the accepted method for performing this task with the '
primary system pressurized.
On November 9, a nuclear instrument channel #3 (intermediate range g monitor) period scram occurred while the reactor was operating at a steady state 10 MW. No actual power transient was indicated on any other nuclear instrumentation. The intermediate range short period rod run in and I scrams are designed to assure protecuon of fuel elements from a continuous startup rod withdrawal accident (HSR. Add. 5, Section 5). If an actual power I
I-8 I
4 I transient had occurred at 10 MW, the three power range instrument scram and rod run-in trips would have occurred before a short period trip could occur, The channel #3 detector drywell and the surrounding pool area were ]
l visually inspected with no unusual indications noted. A front panel check of l channel #3 was completed, no electrical anomalies were indicated, and this
'g parucular problem could not be duplicated. The reactor was subsequently 5 retumed to normal operadon with no further problerns of this nature. 1 g On November 14, both of the operating secondary coolant pumps shut {
down when their substation power supply breaker tripped due to a thermal l overload, his occurred when the motor for secondary pump #2 drew l excessive current and its local breaker failed to trip, which then caused the )
substation breaker to trip. His substation breaker supplies power to the i
I local breaker panels for secondary pump #2 and #3. The primary coolant inlet temperature then rose to the high temperature scram set point of
! 148'F in approximately two minutes, causing a reactor loop high temperature scram. The operators involved in this particular situation did i not respond quickly to the changing temperature because they were I lg occupied with determining the cause of the loss of secondary flow. They have since been instructed that in this situation, the standard operating procedure calls for a manual reduction in reactor power if full secondary
'l flow cannot be restored immediately. It was subsequently determined that the motor for secondary pump #2 had an insuladon breakdown, requiring a complete rewinding. The local supply breaker for this motor was also I replaced.
,g later on November 14, the reactor scrammed due to a momentary loss of site electrical power, This was verified by the University power plant. '
'Ihe reactor was refueled and then returned to normal operation, iI l Major maintenance items for the month included: replacing the ig beryllium reflector: replacing four graphite wedges: wiring the distribu-tion panels for the UPS: replacing valves 550 C & D (pool to primary chec'.
valves): rebuilding the actuators for primary isolation valves 507 A and B; h replacing the detector for nuclear instrument channel #2 (intermediate range monitor); installing bismuth / indium filter in Beamport "F"; replacing a I blown fuse in the local breaker for primary pump 501A: and replacing the auto transfer switch for emergency electrical distribution with a new
" adjustable" switch set at 255 amps.
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l December 1989 g The reactor operated continuously in December with the following
.W exceptions: four shutdowns for scheduled maintenance and refueling; and two unscheduled shutdowns. .
On December 2, a nuclear instrument channel #5 (power range monitor) high power scram occurred. No actual high power condition was I I indicated on any other instrumentation. Channel #5 is one of three power range monitors, all of which will indicate any actual high power condition g and will respond by rod run in or scram. Electronics technicians subsequently discovered, and replaced, a failed trip module in the drawer ,
for this instrument. The reactor was then refueled and returned to normal ;
j operation. ,
On December 11. a rod not in contact with magnet rod run in I. occurred when control blade "B" separated from its magnet during a normal <
reactor startup. The reactor was subcritical when this occurred. The rod l drive amphenol connector was checked and the magnet strength was tested (
sausfactorily. The anvil surface was cleaned and the pull rod was centered within the guide tube. A normal startup was then completed with no further I problems of this nature, g Major maintenance items for this month included: replacing the drawer trip module on channel #5: installing a new breaker on secondary pump #2 controller assembly: Installing a spare trip actuator amplifier in the l rod run-in pos!Uon; and replacing the source range monitor detector.
. January 1990 The reactor operated conunuously in January with the following l- excepdons: six shutdowns for refueling and/or maintenance; and two unscheduled shutdowns.
On January 1, a rod run-in occurred simultaneously with the acknowledgement of an S-1 on limit annunciation. S-1 is an automatically
'g operated hydraulic valve which allows secondary water to bypass the primmy heat exchanger according to temperature demand. An annunciation for this valve occurs when it reaches 20% of its full open or full closed postuon. No unusual or abnormal indications were seen or recorded on any instrumenta-tion at the ume of the rod run-in. After examining all reactor indications
'I l 10 I
I ;
i l and finding no apparent cause, the rod run in was reset and the reactor was returned to normal operation. 'Ihe S-1 annunciation was felt to be coincidental and the rod run in was considered spurious. I On January 31, a manual rod run in was mitiated by the shift ;
g supervisor after observing the channel 5 and 6 power range indications i drifting down below 95% power. The problem was isolated to channel #4, the wide range monitor. This channel is part of the servo mechanism that I automatically maintains reactor power at a preset point. The detector for j this channel was replaced and its drywell was repositioned. The reactor was j g subsequently returned to normal operation.
Major maintenance items for this month included: installing a new air ,
l can in Beamport "A"; installing a new ce' . 'rtube with silicon filters in Beamport "F"; and replacing channel #4 detector (twice).
l February 1990 l The reactor operated continuously in February with the following I exceptions: four shutdowns for refueling and scheduled maintenance; one shutdown for refueling and a Nuclear Regulatory Commission reactor I operator startup examination; and two unscheduled shutdowns.
g On February 1, a reactor scram occurred when a Campus Facility electrician inadvertently opened breaker #2 on the uninterrupuble power supply (UPS) electrical distribution panel #1, while attempting to secure the l breaker for a wall heater in the UPS room. Breaker #2 on the UPS distribution panel #1 supplies power to the UPS. The UPS supplies condluoned power to two distribution panels located in the reactor control '
I room, which, in turn, supply reactor control and instrumentation power.
The opening of breaker #2 and the subsequent momentary loss of control l and instrumentation power caused the scram. The breaker was immediately reset, restoring power. A pre-criucal startup checksheet was completed and the reactor was returned to normal operation upon completion of a hot I- startup. The Campus Facility supervisor of the electrician involved in this incident was informed and the situation was discussed among Reactor l Operations and Campus Facility personnel.
I On February 26. a manual rod run in was inidated by the shift super-visor when no automatic rod run-in occurred subsequent to receiving a channel #4 high power rod run-in annunciation and drawer light. No actual I l-11 I
c - . . - . - . _ . _ . . -_ -_ - . - - _ . . - - _ _ - _.
I i
i I high power condidon was indicated on any instrumentation prior to or at the time of the annunciation. An electronics technician isolated the I
problem to the dual trip unit for channel #4 (wide range monitor) which he !
Ig i
subsequently replaced, ne dual trip unit in channel #4 provides a signal 1 indication for the annunciator and drawer light and a separate signal for the l trip functions (rod run in and scram). The electronics technician I determined that the signal leg for the annunciator and drawer light was f defective and created a spurious annunciator and drawer indication. Before i replacing the dual trip unit, a test of the operability of the rod run in and I scram trip funcuons for channel #4 indicated that the trips operated i
satisfactorily and in their intended manner. After completing a pre-cridcal ll checksheet, which includes a check of all nuclear instrument channels, the reactor was returned to normal operation with no further problems.
]
! Major maintenance items for this month included: performing the j biennial emergency drill with emergency support groups outside the MURR: i l and replacing the dual trip unit in channel #4 (wide range monitor) drawer. l I March 1990 The reactor operated continuously in March with the following 1
1 g excepdons: four shutdowns for refueling and scheduled maintenance: one shutdown for a flux trap sample changeout, here were no unscheduled shutdowns.
I Major maintenance items for this month included: replacing primary l
{
pump 501B with a spare after a leak was detected in the pump shaft seal. l 1
April 1990 I 1I The reactor operated continuously in April with the following l
iI cxceptions: five shutdowns for refueling and scheduled maintenance: and one unscheduled shutdown.
'g On April 8. a manual rod run-in was initiated by the duty operator l when the outer airlock door failed to operate from its closed postuon. The
'g cause was determined to be sucking relay contacts in a door sequence
'E control relay. The contacts were cleaned and exercised satisfactorily and the reactor was refueled and returned to normal operation. The affected g relays for the outer door were replaced on April 9.
1 I
l-12 I -
l 1
,I l
-l Major maintenance items for this month included: replacing the source range detector: replacing the close position relays in the outer personnel airlock door; adding a third securing pin to the threaded end of I all four control blade pull rods: replacing the pool TC - MV/I multivolt transmitter; and replacing the wide range monitor detector.
May1990 h- ne reactor operated continuously in May with the following ,
exceptions: four shutdowns for refueling and scheduled maintenance, g There were no unscheduled shutdowns this month.
Major maintenance items for this month included: replacing the pool
'l delta-temperature summer: completing the annual containment building leak rate test; and pulling Beamport F" centertube back four feet.
June 1990 g ne reactor operated continuously in June with the following exceptions: four shutdowns for refueling and scheduled maintenance:and three unscheduled shutdowns.
On June 6, a reactor scram occurred due to the loss of facility I electrical power during a severe thunderstorm. Site power was restored to the facility in approximately fourteen minutes. At the onset of this electrical outage, the emerency generator started satisfactorily and assumed its l electrical loads. It vras discovered, however, that a main facility exhaust fan (either EF-13 or EF-14) was not operating. EF-13 and EF-14 are redundant.
I 100 horsepower motor driven exhaust fans-either fan will start upon the failure of the (operating) fan. Electronics technicians discovered blown fuses in the control circuits for both fans. He fuses were replaced and an j exhaust fan was started. An emergency generator load test (Compliance Procedure #17) was cerformed and the exhaust fan operated satisfactorily.
The reactor was subsequently refueled and returned to operation.
On June 14, a reactor scram occurred due to the momentary loss of facility electrical power during a severe thunderstorm. Again it was discovered that the ma' saust system was not operating and that the control power fuses for eacn fan were blown. The fuses were replaced and l an exhaust fan was started. An emergency generator load test was I I-13 g 1
I
- l performed satisfactorily and the exhaust fan (s) operated properly. 'Ihe reactor was refueled and returned to operadon.
Exhaust fans EF-13 and EF-14 and their associated controllers are part of a facility ventiladon system upgrade completed in October of 1989.
,g After consultation with the system design engineers and the controller vendor the following modifications were suggested to alleviate the fuse problems: separating the voltage source for control power to each fan so
- l that power for each fan is developed across different phases; increasing the
, amperage rating of the control fuses from 3 amps to 6.25 amps for each fan
'g which would also require replacement of the 480V/120V transformer a currently rated at 250 VA with one rated at 750 VA. The first modification has been completed. The second modifleation has been completed on l EF-13 and will be completed on EF-14 when a second control power transformer arrives. ' A Licensee Event Report was sent to the Director of ~
lg Nuclear Reactor Regulation on July 6,1990 detailing these unscheduled Ig shutdown events and the subsequent corrective actions, g On June 18, a rod not in contact with magnet rod run-in occurred when control blade "A" disengaged from its magnet during a normal startup. ,
The reactor was subcritical at the time this occurred. The anvil and magnet !
were checked and cleaned and associated wiring was inspected. Rod alignment to the offset tube was checked. No anomalies were noted. A normal reactor startup was then completed with no further problems of this type.
Major maintenance items for this month included: removing the l silicon filter centertube from Beamport "F" and installing the iridium /
bismuth filter centertube: replacing the control power fuses on exhaust ~.w EF-13 and EF-14; replacing the reactor temperature (Tc ) multivolt I. transmitter: repairing worn leads on the Th RTD for the in-pool heat exchanger; and replacing the control circuit transformer on EF-13.
i I ;
I I
l-14 I
- -. - -- .- -. - _ . . .- - ~_- _ __ _ ..
I i I i SECTION II )
( CHANGES 'IO 'IHE STANDARD OPERA'I1NG PROCEDURES 2nd Edition, Effective Date: 5/02/89 (Revisions #1 through #24 to the October 1981 printing 1 5 were incorporated.)
{
1 July 1989 through 30 June 1990 t
8 As required by the MURR Technical Specificadons, the Reactor i
i Manager reviewed and approved the following:
Revision No.1, dated 8/14/89 Revision No.2, dated 1/16/90 t
CHAtMES 'IO THE M'URR SITE EMERGENCY PROCEDURES AND )
' FACIIIIY EMERGENCY PROCEDURES (dated January 1985, and revised May 13, 1988) j NOTE: New manual printed May 13,1988 1 July 1989 through 30 June 1990
,3- Je required by the MURR Technical Specifications, the Reactor 3- Maager reviewed and approved the following:
Revision No. 3, dated 9/18/89 I- Revision No. 4, dated 12/8/89 Revision No. 5. dated 3/22/90 NOTE: SEP-7, page 2: Social Security Numbers have been omitted per 1 request by Nuclear Regulatory Commission, Al Adams,9/87.
The revisions to the Standard Operating Procedures and Emergency ,
I 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 ( l ).
I I
LI '
Il-1 I
I I STANDARD OPERNI1NG PROCEDURES NANUAL 2nd Edition, Effective Date: 5/02/89 (Revisions #1 through #24 to the October 1981 printing :
were incorporated.)
Revision Number 1 -;
Revision Date: August 14,1989 Page Number - :,
, SOP /117 l SOP /I-18 Reset f SOP /I 19 Reset l SOP /I-20 Reset
! SOP /I-21 SOP /I-22 Reset SOP /II-8 SOP /II-13 l SOP /III-9 l SOP /V-3 i l SOP /VI-9 l SOP /VIII-33 I i
SOP /VIII-35 GOP/VIII-37 REP-9-3 h
REP 14-1 L REP-20-2 SOP /A-8a Retyped SOP /A 8b ,
SOP /A-8c LI SOP /A-8d Retyped r
SOP /A-12c II 11- 2 I
I :
l If direct communications are lost or if one of the above reports is not acknowledged, reactor power will be main- l tained at a steady level until the problem is corrected.
The Health Physics Technician will make his final report ,
to the Control Room after a complete survey is conducted at ;
the-desired power level. ;
1.4.4 Normal Operation l
.i A. Normal power level will be 9.90 to 10.00 MW as indicated by j l
the total power meter.
B. The control roon shall be occupied by at least one licensed B operator during steady state operation of the reactor. A ]
W j second licensed operator must be at a facility location ] ,
I where communication with the control room can be maintained. ]
C. Prior to assuming control of the reactor, the oncoming oper- l ator will read the control room log book and shall be briefed on current operation.
D. During shift operation, the Shift Supervisor for the new shift will review the log book and be briefed on current L operations by the crew he is to relieve. Upon completion of ,
the log book review, the Shift Supervisor will note the same i
in the log book. !
E. A complete set of Nuclear data will be taken once an hour '
L - during steady state operation.
F. A complete set of Process data will be taken every two (2)
L hours during steady state operation.
G. During routine operation, a routine patrol of the facility
)3' l
will be made every four (4) hours according to an approved E Routine Patrol Checksheet.
H. Normally, the calorimetric determination of the power level can be read directly from the digital readout and entereo in ,
the Nuclear Process Data. The cause of any difference between the primary and secondary calorimetric calculations which exceeds 5% (0.5 MW) during steady state full power
'g' u! operation should be determined. The primary power
' I- Rev. 8/14189 App'd ,l d S0P/1-9
,I
I i I D. All experimenters will be required to complete an indoctri-nation training course on the relationship between-his i
I I experiment and reactor operations, emergency procedures, and radiation safety.
- E. All reactor users shall complete a Reactor Otilization j Request form. This request must be reviewed and approved by j the Reactor Manager. )
F. If an experiment appears to involve new or unevaluated ;
i hazards, a review of the proposed experiment by the Reactor Advisory Comittee may be requested by the Reactor Manager.
G. The Reactor Manager may require, as deemed necessary for safe operation, that experimental data or operating instructions ]
l be on file with the reactor operations organization. )
H. All changes in beamport experiments, and intentional flood- !
ing or draining of. the beam tubes to be performed with the reactor at power will be done only af ter written approval is l g i E. initially obtained from the Reactor Manager and after a pro-cedure for so doing has been established. Whenever practi- )
cal, beamports shall be flooded or drained only after the l reactor has been shutdown for at least 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 1. The insertion and removal of experiments in the center test hole position will be done with the reactor shutdown.
l 1.4.7 Modification Records ]
l The purpose of the modification record system is to document ]
-f. changes to reactor license related systems as described in ]
Hazards Summary. ]
l These records must include a written safety evaluation which ] ,
provides the bases for the determination that the change does ]
not involve an unreviewed safety question as defined in 10 CFR ]
50.59. ]
I lf Rev. 8/14/89 Aop'd ([M SOP /1-12 g-
I 1.4.7.1- Procedure for Initiating and Processing Modification Records ]
NOTE: Please contact the Modification Records Manager before
- _I ]
initiating any modification packages. The original copy 3 of each modification package will be kept separate from ]
the available copies in the file cabinet. The Modifica- ]
tion Records Manager will update the originals as ]
necessary.
]
A. When initiating a modification, first obtain a blank modifi- 3
'I' cation package from the file.
DO NOT assign a number to the modification package or log it ]
]
L in the notebook at this time! .]
- 8. Complete a preliminary modification proposal. Even though ]
I it is preliminary, try to make it as complete as possible.
]
If the proposal req? ires a drawing or print, include it if ]'
available. For assistance, see Modification Recnrds Manager. ]
(1) When completing the safety evaluation and unreviewed ]
safety question sections, list the Hazards Summary ]
Report (HSR) and the Technical Specification sections ]
that apply or the reason they are not affected. ]
C. When the preliminary proposal is written, submit it for ]
crew review and evaluation. This is where comments and ]
suggestions are solicited. Suggested changes can then be .]
incorporated into the final draft. ]
D. After the proposal has been reviewed and necessary changes -] l I have been made, submit the final draft to the Reactor Manager ]
for review. ]
(1) The Reactor Safety Analysis (pg. 3) should be put in ]
final form. If the modification involves a change to ]
the Facility as defined in the HSR, include suggested ]
revisions to the HSR, referencing the applicable ]
section of the HSR. ]
~
l If it does not affect the HSR, then outline the basis ]
for the decision. ]
Rev. 8/14/89 App'd , M S0P/I-13 ]
l n
I i
(2) The Reactor Safety Evaluation (pg. 4) should be put in ]
final form at this time if it involves changes to the ')
Technical Specifications or an unreviewed safety 3 l g estion. ]
E. If the modification is determined by the Reactor Manager to ]
]
'be an improvement, it will then be assigned a number and ] 1 routed for final crew evaluation. If not, it will not be ]
I assigned a number and it will be cancelled at this time.
F. After the f M 1 crew review, it will be ready or approval
]
']
)
)
by the Reactor Manager and will be reviewed by the Safety ]
Subcommittee and/or Reactor Advisory Committee as necessary. ] ;
G. The modification will then be completed as soon as possible, ] l l
including updating any necessary prints, Standard Operating ] j
]
I H.
Procedures, Compliance Checks or Preventive Maintenances.
When these are completed, the modification package will be ]
routed by the Modification Records Manager to the Reactor ] 1 Manager and signed off as completed. ]
l 1.4.8 Radiation Work Permit ]
A Radiation Work Permit will be completed by the Job Supervisor and Health Physics prior to conducting any work which in the I opinion of the Shift Supervisor or.the Health Physics Group l L .
involves significant potential for exposure of personnel to l l
radiation or the spreading or release of airborne or surface l \
L contamination. l A copy of this form is included in the Appendix of the SOP. ]
The form is used as follows:
l 1
l l
Rev. '/14/89 App'd Mk S0P/1-14 ]
1I
I A. The' top portion of the RWP will be prepared by the Job .
Supervisor and Health Physics. The time and date spaces i should contain the supervisor's estimate cf the duration of the job (0800-2400 September 2, 1988; 0000-2400 September 10-12,1988,etc.). The Job Supervisor should ;
be as specific as possible in describing the job. This will aid Health Physics in determining the protective measures .necessary. "
Health Physics will assign a number to the RWP and conduct I
B.
the necessary surveys and determine the protective measures i necessary for the job. Health Physics will coniplete the remainder of the RWP indicating the survey results and the protective measures required. Health Physics will then sign and date the RWP, and obtain approval signature from the Job Supervisor and information signature from the Shift j i
Supervisor, if appropriate. I C. The Job Supervisor will provide Health Physics with the names of personnel expected to work under the RWP prior to the start of the job. Each person performing work under the RWP 1 shall be informed of the required radiation controls and ,
shall acknowledge being informed by signing on their '
designated signature line. Health Physics wi'l have avail- l able the approved RWP at the job site for ready reference by the personnel doing the work. ,
D. When the job has been completed and the job site has been cleaned up and decontaminated, the Job Supervisor will deliver the RWP to Health Physics. Health Physics will verify that the job site is clean and decontaminated, and will record i estimated dose for each person involved. Health Physics will tt :ninate the RWP and maintain it in an RWP file. Any person who signed approval or any supervisor can terminate the RWP I by signing the termination block. Health Physics must be notified if the RWP is terminated by someone other than a member of the Health Physics Group.
Rev. 8/14/89_ App'd @DM SOP /I-15 )
Reset I <
I l
1.4.9 Radiation Safety ]
The Shif t Supervisor is directly responsible for the overall safety of personnel on his shift and indirectly responsible for ,
all peiaonnel whose safety may be affected by activities con-ducted under his supervision. Radiation safety is a very im-portant 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 person-I nel, but rather that the Shift Supervisor is responsible to insure that through coordination with the Health Physics person-i nel, adequate protection is provided for evolutions conducted lI i
during his shift.
3
'm 1.4.10 Physical Protection of Special Nuclear Materials ]
g 1 In accordance with 10 CFR 73, special requirements must be met in safeguarding Special Nuclear Material. The safeguards provided I and the procedures applicable to maintaining the security of Sp.cial Nuclear Materials are contained in the facility-Security Plan and Security Procedures. e 1.4.11 Equipment Tagout Procedure ]
1.4.11.1 Purpose ]
The purpose of the tagout system is to prevent injury to person- ]
nel and damage to equipment. ]
1.4.11.2 Types of Tags ]
A. Red Tags -]
Red tags will be used to identify .quipment which, if oper- ]
ated could present a hazard to pe sonnel. The tag will also ]
contain information as to the potutial hazard. ]
4 Rev. 341.4]89 App'd h SOP /I-16 ]
I
c I
B. Yellow-Tags -]
Yellow tags will be used to identify equipment which, if ] H operated, could present a'nazard to equipment. The' tag ] .
L will alwo contain information as to the potential hazard. -] 1 1
-1.4.11.3 Equipment Tagout ]
A. Tagout Log ]
- 1. The tagout log will be maintained in the control room ] l by the on duty shif t supervisor. -] ;
I 2. The_tagout log will be mada up of four sections:
- a. Tagout Instructio1s
]
]-
I J
- b. Index ]*
l k c. Active Taput Sheets ]
- d. Cleared Tegout Sheets -].
B. Performing a Tagout ]'
- 1. A tagout will be act -@shed by a licensed operator ] (
L - only. ]
- 2. The on duty shift supervisor must approve the hanging ]
L emgl or removal of any tag. ']
- 3. In the event that a tag is uissing, the shift Luper- ]- 5 l:
visor will be informed immediately. . The missing tag ]
will be cleared from the tagout sheet and a new tag ] .i issued. ]
C. Tagout Audit ]
- 1. A tagout audit will be performed at least monthly. ]
l
- 2. The tagout audit will consist of verifying that the .] ,
l 1
index includes each active tagout and also that each ]
k l tag for the active tagouts is still in place and ]-
correct. ]
b 3. Upon completion of the audit, the person completing ]
the audit will sign and date the tagout audit sheet.- ] ..
l 4. Upon completion of the audit, all the tagouts in the ] I l inactive section of the tagout log may be discarded. ]
l Rev. 8/14/89 App'd M S0P/I-17 ]
I Table 111' l' Normal Reactor Operating Ranges Parameters Normai Operating Range Units t __
- 1. Thermal Power, 5 ftW Operation 5 5% t1W
- 2. Thermal Power,10 llW Operation 11W 10'+g%' %
a h
- 3. Primary Coolant Flow, 5 MW Operation 1850 t 50 gpm
) Primary Coolant Flow,10 MW Operation 3700 50 gpm l_
- 4. Reactor Cutlet Coolant Temperature 136 "F
- 5. Reactor inlet Coolant Temperature 120 *F. !
- 6. Pressuriz0r Pressure 67 + 3 psig
- 7. Pressurizer Level CENTERLINE +~4 to -8 inches
- 8. Pool Coolant Flow, 5 ItW Operation 600
- 100 gpm h
L33 l. Pool Coolant Flow, 10 MW Operation 1200 100 gpm
- 9. Pool Outlet Temperature (Hot Leg) 105 F 10.-Pool Level 29' 7" t 3" feet-inches
- 11. Resistivity, Outlet # D: 300 >500K ohms-cm p 12. S-1 Temperature Demand Set 120 F 1
ll t
Reset
Table IV flominal Values of Trip Settings for- Alarm, Run-in and Scram Conditions for 10t1W Operation Scfaa ' Run-In- Alarm Units
- 1. Short Period 9 11 sec-
- 2. Low Count Rate -- --
<1.0 cps
- 3. liigh Poder 119 114 --
7, full power
, 4. RC Inlet ~ Temp 148 --
140 'F
- 5. RC Outlet Temp 168 --
160- *F-
- 6. RC System Low Flow 11725 --
1800 gpm 10 11W Operation
- 7. Ileat Exchanger Low 31675 -- --
gpm
& (DPS 928A/B)
- 8. Rx System Low Press 26 3 -- --
psig Switch PS 944A/B
> 9. Core Low d , 10 flW 33300 -- --
gpm l
- 10. Low Pressurizer Level 13 below C.L. --
10 below C.L. inch
- 11. liigh Pressurizer Water 14 above C.L.
inch Level 1 Alarm and Scran received from either . loop 2 Pressurizer' pressure with nonnal system flow 39 Corresponding to this flow value Rev. g4/33__ App'd 50P/I-19 ]
i Reset '
f
., 'i t
, , _ . - _ _ .._. , .A., .
W
-W.-m W W
.M M~ . W
M' 'Wf (M: ~M : M- M( LM: ML M=: M 3.
LTABLE IV-(continued)
~
Scram' Run-In Alarm Units
- 12. Low Pressurizer Press- 63 --
65 psig
- 13. Hi Pressurizer Press 78 --
75 .psig
- 14. Pool Low Flow, 10 fM 490 --
530 gpm
- 15. Pool lli Temp -- --
115 *F
- 16. Low Pri Demin Flow -- --
42.5 gpm
- 17. Low Pool Demin Flow -- --
-42.5 gpm
- 18. Bldg Air Plenum Hi 10 x normal oper- -- --
cir/hr -
Activity ating background
- 19. Reactor Bridge 10 x normal oper- -- --
mr/hr ating background
- 20. RC Hi Conductivity -- --
2.0 pahos
- 21. PC lli Conductivity -- --
2.0 unhos
- 22. Hi Ref1 d , 10 IN 7.0 -- --
psi
- 23. Low Refl @ , 10 IM 3.0 -- --
psi
- 24. Low N2 System Press -- --
115 psig
- 25. Low Seal Trench Level -- --
5 feet
- 26. Hi/Lo Level in T-300 -- --
6200/2500 gal.
- 27. Hi/Lo Level in T-301 -- --
6000/<10G gal Rev. 8/14/89_ App'd M ' SOP /I-20-]
. Reset p.
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ _ = _ _ _ _ _____: -a - . . . . .~ :
WW W
~g' g*
~
W z g - ig._ Ig ; i gi j g 13.~ E g'=hmi g' g - ig i g) j.3 g -
I. .
~
l . TABLE IV (continued)
Scram ~ Run-In Alarm Units
- 28. Fission Product -- --
00 cps 11onitor Hi Activity 5
see below cpm
- 29. Off-Gas Hi Activity --
- 30. Secondary Coolant. -- --
-10 cps Hi Activity
- 31. Anti-Siphon Line Hi ---
>6 --
inches
~
Level (above valves)
- 32. Pool Level Low >24' >28' --
feet ]
l '33. Reg Blade --
<10% or '<20% or >60% % withdrawn l bottomed i
! 34. Vent Tank Low Level --
7-11 --
inches' (below C)
- 35. Secondary Coolant -- --
<1800 gpm Low Flow
- 36. Ch 4, 5, or 6 Downscale -- --
<95 % full-scale
- 37. Valve 546 A or B -- --
off closed
- 38. Valve 509 off open -- --
- 39. Valve 547 -- --
off open
- 40. Valves 507 A/B off'open~ -
closed with P501 on
- 41. Valve 5-1 -- --
90% open or --
90% shut
'5 This setpoint1is' determined by the semiannual calibration.
Rev. 8/14/89_ App'd. MI%Y\
50P/1-21 ] <
i q
-- - ;,. = - . -
l~J M ; W .: M W- : 'W :' W W W
~
~i m : W f M .- M ? M '. W 'i .
- M -
l
( -
t TABLE IV (continued)'
~
Scran. Run-In Alarm" Units l
6
- 42. Fluclear Instrument inoperative --
~
anomaly --
- 43. Anti-Siphon System --- --
30 psig Pressure. Law Anti-Siphon System -- --
44 psig Pressure liigh
- 44. Thermal Colunn Door. -- -. .open --
- 45. Truck Entry --
door seal -- --
deflated
- 46. Evacuation or manual / auto --
manual / auto --
- i Isolation 47 Rx System Low 36 3 -- --
psig Pressure (PT-943)-
l l
l I
l 1 l-3 6
Pressurizer Pressure with nccmal system flow Any channel' will scram on fil. Inoperative except_ SRil Rev. 8/14/89 App'd ~ EM 50P/I-22~ ]-
Reset
. E:
_w- _.~ . . , . .
I 11'.1.6 Reactor Shutdown Procedure I A ~. The procedure for a routine reactor shutdown requires only that the manJai rod run-in circuit be activated. However, prior to shutting down the reactor:
- 1. Turn on the source' range recorder and time and date-the chart.
- 2. Insert the fission chamber until a count'of approximately 105 cps is obtained on the SRM recorder.
- 3. Place the IRM recorder in fast speed and time and date
- the chart.
- 4. Take a set of nuclear and process data.
B. Depress the manual rod run-in button on the control console.
' Enter the time of shutdown in the log book.
C. Follow the reactor power decrease by changing the range-selector switch so as to keep channel WRM-4 on scale.
D. Complete the Reactor Shutdown Checksheet if the control room '. ]
.is to be left unattended for an ' extended period. ]
E. Ensure that the-primary and pool systems-are shutdown as per S0P IV.2 and V.2 respectively if the control room-is left
- unattended for an extended period of time.
11.1.7 Reductions in Power
- The procedure for reductions in power to perform short evolutions-(<- 45 minutes) such as Room 114 entry. shall be as follows
A. Turn on source range recorder and time and date chart.
B. Inset fission chamber until a count of approximately 10s cps is obtained on SRM recorder.
C. Place IRM recorder to fast speed and time and date the chart.
. Take a set of nuclear and process data, i
E. Depress the manual rod run-in button on the control console, i F. Drive control rods in 3" or to a height of 21" withdrawn, whichever corresponds to a lower rod height.
G. After evolution is completed, recover powt:r following pro-cedure for hot startup (11.1.2).
H. If the evolution for which the reduction in power b made exceeds or appears that it will axtend past 45 mirates, shut down the reactor following proceduce 11.1.6.
Rev.8/14/89_ App'd D SOP /Il-8 I .
i ag F. During' the process of' actually removing the offset, the core
's neutron levels will be continually monitored using the- t fission-pulse channel SRM-1. The reactor control room may i be unattended during the removal operation. All reactor l systems will be shutdown.
'. t 11.3.2 Caution Should be -Taken During Removal as Follows:
A. Due to the potentially'high radiation level produced by the activated blade, place the bridge arils to the upscale posi- -
tion to prevent a building isolation alarm while the offset is being handled near the surface of the. water.
gg B. Extreue caution should be used during Step II.3.3.L so that undue stress is not placed on offset mechanism'while breaking it loose from. guide pins. Also, extreme caution should be
) used while maneuvering the offset mechanism away from the pressure vessel.
11'.3.3 - The Detailed Procedure for Removal is as Follows:
A. Electrically disconnect the rod drive mechanism. ,
B. Remove the four bolts at the base of the rod drive mechanism and with either the rod magnet fully inserted or withdrawn, l remove the rod drive mechanism.
C. Remove the four bolts at the base to the rod' drive shaft housing assembly.
D. Uncouple the amphenol connections to the drop timer / rod bottom ]
I .
photocells. Ilark position of photocell housing if it is to be ]
removed. ]
E. Remove the "V" clamp attaching the upper housing to the L bridge floor plate.
Lg: F. The upper housing unit may now be lifced free. ilark the 3 upper housing when more than one mechenism is to be removed.
l G. Loosen the bolts on the lower housing tiracket and remove same.
L .:
Rev. 8/14/89_ App'd M
. SOP /Il-13
.i n ,
~
. G '. Verify that the particulate recorder and gas recorder. pens <
inaicate 3500 cpm .* 10% and that the stack monitor high-
! . activity annunciation is received. Also verify-that the -)
- i local meter reads 3600 cpm + 10%. ]: l H. Return the particulate mode switch to "0P" position.-
- 1. Press the " reset" button until the particulate meter and !
. recorder readings return to normal; do not drive it to the l downscale position. Reset the annunciator. I J. Test the low flow alarm in the control room by securing the
/ blower.
Return the blower switch to "on", verify "high" and " low" K.
alarms cleared.
t 111.8 Area Radiation lionitoring System The. area'ra. stion monitoring system will be in operation con-tinuously and is to be turned off only during maintenance on the system. When-handling samples or during maintenance place the '
Bridge Upscale Switch in the upscale position. insure Bridge
(
' U Upscale Switch is returned to normal position after handling
(. samples. q The station trip points shall be set as follows: i Station 1-BP South Wall 2 X acceptable background
[ i Station 2-BP West Wall 2 X acceptable background
[ Station 3-BP North Wall 2 X acceptable background ,
Station 4-Fuel Vault 10 mr/hr Station'6-Room 114 2 X normal operating background Station 7-Reactor Exhaust 1 mr/hr or 10 X normal opera-Plenum ting background :
.' Station 8-Reactor Bridge 50 mr/hr or 10 X normal opera-o ting background 3
Station 9-Reactor Bridge Backup 1 K - 10 K mr/hr At least once per month the system will be checked according to the following procedure. ;
l u
Rev. 8/14/89 App'd MWN\. SOP /lil-9 3 -
~ ~ - -
II V.2 Pool: System Shutdown Procedure The pool . cooling system should remain in operation' for a short ;l g
i V.2.1
- period of' time .(5 minutes minimum) after a normal reactor shut-3 down in order to remove core decay heat from the reflector and j experimental facility. The procedure for attaining a normal.
pool system shutdown mode is as follows:
A. Place master switch I'S1 in test.
B. Turn off P508A/B using the control switches in the control' ,
room. To minimize check valve slam, secure both pumps simultaneously.
C. Verify that valve V509 closes and cleanup pump P5138 shuts ]'
off automatically. ] .!
]
~
D. Turn cleanup pump P5138 switch to off.
E.. Place V509 in the manual / closed position.
F. Verify that all the valve position indicating lights are
- operating. If not, replace the appropriate light bulb. If this does not clear the malfunction, determine the cause and make repairs prior to any reactor start up.
I.
NOTE: The following steps are at Shift Supervisor's discretion..
G. Turn off the pool flow and temperature recorders.
'H. Secure power to P508 A/B. ;
V.3 Partial Pool Filling Procedures (Pool at Refuet 1.evel or ' Above)
.V.3.1 To increase the water level in the pool with demineralized water j
- f rom T301 or T300, one of the two following procedures can be-
- used; hcwever, all water in T301 shuuld be used first.
.- A. Filling may be accomplished with the skimmer system (Section VII.S.3) with or without the skimmer pump operating and the reactor either cperating or shutdown. Required operational pool makeup will be accomplished in this manner:
- 1. Check capacities of tanks T300 and T301 and check proper valve lineup.
- 2. Observe the pool level and check that the skimmer pump is secured.
Rev. M431 AppN Whd SOP /V-3 I .
- VI'.6.6 Secondary Silt, Algae and Mud Control Silt and mud-buildup is controlled by the feeding of- a chemical, q silt dispersant to the cooling tower basin. The. dispersant is-g
- WI added to ensure solids remain suspended a sufficient amount of time to allow the secondary blowdown to remove them from the
- /
j
, system. This reduces secondary conductivity and minimizes -the tuildup of silt in low flow areas, a fouling condition. '
addition of two microbiocides/algaecides. The addition-fre-quency is determined by weather conditions and reactor oper- i stions.
VI.7 Secondary System Operation on Maintenance / Refueling Days ]
p This section is to be used to ensure sufficient cooling is ]. ]
provided to' the LiBr air conditioning unit when a secondary ' -] ,
j pump is required to maintain primary and/or pool system ]
.g ]
E. temperatures. I A. Place P-4 in hani 30 it will run continuously. ]
NOTE: Fans may need to be run also to maintain proper cold ] 4 l :
i deck temperatures. ]
B. Start either P-1, P-2, or P-3 and run as necessary to main- ]
l tain primary and pool system temperatures. ]
C. When the reactor is placed in automatic at 10 MW, return ]
P-4 to automatic. ]
I! .
I LI Rev. gL)3fAL pp'd k SOP /VI-9 L
I
- f' :
-Vill.4.6 Operating Procedures {for Beamport'"F" The following procedures shall be used for operation of Beamport
- B "F". - All valve and tube changes shall be made by Reactor _ Opera-
, tions personnel. 11ajor shielding movements and all center tube adjustments or changes. shall be coordinated with Health Physics -
and Reactor Operations personnel. A copy _of this procedure
! shall be posted near Beamprt "F" and a copy shall be put in the L ! .Beamport "F" log book.
g f 3E CAUTIO!!S:
. Insure center tube is not le't fully inserted; allow d at least 1/4 inch for t.hermal expansion.
e Af ter the center tube is in3erted, verify the drain and vent valves are shut.
e To prevent a partially filled ocam tube leaving a crack for radiation, be sure the vent tank has water in-it.
e To-limit handling of a vary radioactive filter tube,
~
pull the tube back four feet and let'it decay for > 2 days before withdrawing it. Have Health Physics coverage.
- To limit tritium release, limit leakage of water.
-* To prevent excessive personnel exposure, make sure -!
filter parts are in tube and pushed forward to reactor l end of filter tube. Apply vacuum slowly so that filter
. parts are not sucked back. Have Health physics coverage on startup.
s Af ter startup, check radiation survey readings against ] !
previous readings with similar filters.
- llake it a habit to stay out of beams, whether they are 5 open or " closed".
NOTE: The experimental can may be flooded or drained only when the reactor is shut down.
I h Rev. 8/14/,39, App'd M SOP / Vill-33 l
B
!I 1
- % .8. The vacuum pump shall be hooked up and
- started before-a the reactor is taken critical. The vacuum should be. ,
applied slowly so that suction-will not pull back the -
LIL filter parts.
- 9. A beamport radiation survey shall be completed af ter-
) the reactor is started up at 10 MW.
B. Adjustments to Beamport "F" Center Tube j
l The center tube shall only be adjusted ~with the reactor ,
suocriticali- Adjustments include changing the distance the -
center tube is from the core-and pul. ling or. adding parts. !
A.:
from the center tube.
- 1. - Take the reactor subcritical before adjusting the l center tube.
- 2. If-the center tube is moved, insure it is not closer l than 1/4 inch from being . fully inserted.
- 3. After adjustments are made and vacuum restored, return -
reactor to normal operations, and perform a Beamport "F" radiation survey.
g C. Removing Center Tube from Beamoort "F" i The conter tubes may be very activated. Therefore, close i3
-l Health Physics assistance is required. Minimize the '
number of personnel in Beamports "D",. "E", and' "F" areas while transferring the center tube, 1
- 1. The center tube should be allowed to decay for > .2 days ]
before moving from the beamport. ]
- 2. Af ter loosening the packing nut, pull the center tube l .
back slowly, drying the center tube as it is'being ] '
-withdrawn.
l
- 3. When the center tube is one to two feet from being fully- l withdrawn, attempt to gently close the ball valve (be careful not to score the valve or the center tube).
T'
'I Rev. 8/14]89 App'd _M SOP /VIII-35 1.
l u
. . m.
BRUfeff06 40 500073446 mm em ime -
== um == = == um == == == .
uma
~ .,
~
3 em o 3 me i
g l BP-FG 4 e
SURGE COLLIMATOR STW
( w vavo W -;
i TA Aix ' 3*, g- g 4 f b$ :l
,o (t:: - t
- E :!
57 l sootoGICAL' SHIELO
[ h
& /
, BP-Fi
!.i i0 1:
@*.. es r.
'[E:![{h5I5lffbfhf:h5
...: *:: s.,
e,efr .. ,g.. . :.... ,, .
- .s.a..,.,,g.
. 4,
,.-e. .b
- &.~
9 :f.v}. -:.
- / .( (, o _i =
l! > O
*Om w :
HOUSING VEAIT : $. .L ,3 4 Q*y
. , ...... .... _. }, y gd,5.
?
f.\<, f .. 7 u h L y ** 7
- a. ,l,.K >
. .-. .:-: . . . . .. ,, , tn
- ~... '; . , . .... . . .
i:-1.*:
' %..W. . .: .' .Q.
*. f(l' .o }u p .
.'. . . .'. o .e . . . :~ -
' .._. / ( llp e -< m 1 VACUU M y -,
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- : : - ' ' , ,. o .~.,,,,,,,,o
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. , . . , , , r th !
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..o......... . .. .. ....=. ..o..... ...........-
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*I 6 * -
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.* * .s.Y :,-
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* ', .?:$!.0 b-:l.'g<. ***. .;-
* *$$i .kOG;'. C-fl;f.{': *'
_GO , v g REMOTE
- OPERATED DWAIM VALVE 25
- 4
$s qq '
E g; et l
, .w -
[ 2 j . t . t t
. . tt
JMem
'I j
'l W B. SUSTAINED LOSS OF ELECTRICAL POWER j IMMEDIATE ACTIONS: L 1. Check the reactor shutdown.
- 2. Turn 0FF all pump and cooling tower fan switches.
L 1 3. Place all valve controls in their normal shutdown position and manual mode.
- 4. Trip the master supply breaker on substation "B". -] .
SUBSEQUENT' ACTIONS: (. 1. Notify the Shift Supervisor. l
- 2. Check emergency generat r and its loads for proper operation.
L o 3. Check gas tank level a sd project remaining run t .ce of E. G. t- 1
- 4. Determine cause of electrical power loss. -
1 l 5. If a rabbit is'in the reactor, transfer P-tube blower to l l l-1 3 emergency power and return the rabbit.
]
- 6. Trip the supply breakers for MCC-1, MCC-2A and MCC-28 in cooling tower.
- 7. Make' console log entry and fill out UNSCHEDULED SHUTDOWN report.
REC 0VERY ACTIONS:
- 1. Check all three phases.on each substation for proper voltages..
- 2. When starting the system, closely monitor any equipment known l- to be running at the time of the electrical power loss.
- 3. A Full Power Startup Checksheet shall be performed prior to ,
starting up the reactor. C. M0MENTARY LOSS OF ELECTRICAL POWER: (only a reactor scram occurred) I= 1. Notify the Shif t Supervisor.
- 2. Verify momentary loss of electrical power with power plant.
- 3. The reactor may be operated af ter performing a Reactor Short Fona Precritical Checksheet.
- 4. Make console log entry and fill out UNSCHEDULED SHUTDOWN report.
Rey, 8/14/89 App'd M REP-9-3 I e
REP-14 ! LOSS OF POOL FLOW DURING REACTOR OPERATION IF pool flow rate drops below 490 gpm in either loop w!thout generating -] an automatic scram, the reactor operator shall:
-l I l 1
IMMEDIATE ACTIGNS:
- 1. Scram the reactor.
1 ,g! 2. ' Shut down the pool sys, ;m, leaving the priraary &nd secondary on . !g- the line. h SUBSEQUENT ACTIONS: E 1. Notify the Shift Supervisor. ' 1g. l 2.: Determine the cause of pool flow loss and correct it before restarting the reactor. I
- 3. - Make console log entry and fill out UNSCHEDULED SHUTOOWN reports.
l I! 1 l l , l i 1 l i Rev. 8/14/89 App'd M REP-14-1 l
'3 IF the source of.-activity is determined to be fission products, a I 5.
Turther reduction of primary flow to approximately 500gpm is'necessary to reduce plate erosion. To ac;omplish this: i WARNING: DO NOT ENTER ROOM 114 UNTIL ABSOLUTELY ] ; NECESSARY. A HEALTH PHYSICS' SURVEY IS ] VITAL PRIOR-TO ENTRY. ] ,
- a. Fully open the bypass valve (538A or 538B) around the pump that is running. !
- b. Throttle valves 540 A and B. !
- 6. Clean up contaminated s stems by:
- a. Leaving the primary cooling and primary cleanup loops in operation to clean up the primary system.
WARNING: THE RADIATION LEVELS IN THE DEMINERALIZER ROOMS MAY BE EXTREMELY HIGH.
- 7. When equipment and personnel are ready to identify the leaking fuel element. the primary systems should be shut down as per S0P IV.2.
- 8. Identify the leaking or ruptured fuel element. The fuel element which is leaking fission products must be accurately . identified and placed ,
in safe storage before the remaining intact elements may be utilized. l This will be accomplished in the following manner: j
- a. Have the Health Physics personnel move the portable gaseous and particulate monitors to the reactor bridge for continuous
- moni toring. Health Physics personnel will be present.
- b. Move each element from the core to the "X" or "Y" basket.
l
- c. Draw a grab sample from above each element and give to the -l laboratory group for analysis. If one of-these samples I indicates fission products present, this element will be !
inspected first.
- d. Move the fuel element from one of the baskets to the fuel l 1 spection rig for visual inspection.
- e. The Reactor Manager will determine the disposition of the-leaking fuel element (s).
1 l l: Rev. 8/14/89 App'd i h REP-20-2 j ! l !- l
?
'5 REACTOR ROUTINE PATROL Date:
I 2
- 1. : Time of start of patrol
~
-2.. Time 4.id date all charts
- 3. Check ARMS trip settings L 4. Visual check of entire pool
- 5. Anti-siphon tank pressure 36 psig i 3 psi
- 6. North iso door seal pressure 18-28 psig
.7 . South iso door seal pressure 18-28 psig
- 8. 5th level backup doors Open l
- 9. 5th level detector reading 0-3.5 mr/hr
- 10. 5th level trip point set 3.5 mr/hr
- 11. 16" iso v1v A air pressure 45-55 psig
> 90 psig 3
- 12. 16" iso viv B air pressure L
- 13. Emerg compress on standby ,cg,gg gen, L
yg
- 14. Containment hot sump pumps Operable
- 15. Door 101 seal pressure 18-28 psig
- 16. BP floor Conditions normal.
1
- 17. Fuel vault Locked
! S, i n 18. Inner airlock door seal press. 18-28 psig
- 19. Outer airlock door seal press. 18-28 psig
- 20. Cold-deck temperature 51 2 4*F 3
- 21. T-300 level >-2000 gal.
- 22. T-301 level < 6000 gal.
- 23. ' Labyrinth sump Level < Alarm Pt.
On ene first. routine patrol of the day or tne first patrol after a startup, drain all water from the anti-siphon system. if draining causes the pressure to drop signifi-cantly, return to the middle of the band (36 psig) and record the pressure here, if a condition or reading is normal, enter a "/" (for conditions) or the reading in the applicable box. If the condition is abnormal, enter the condition or reading and circle it. Explain all abnormal conditions or readings in the REttARKS on page 3. i Rev. 8/14/89 App'd (dtWA S0P/A-Sa Retyped Only I l
ACTOR ROUTlHE PATROL ]
- 24. R0 unit power ON
- 25. R0-unit temperature 24-28'C or Standby 26..-'R0 unit pressure g-gpsigor
- 27. EG Rm. t Thermostat > 50*F
,Batt.
EG OP check Sun.Automids.'? switch to Temp > 40*F (Gas l sight glass )
- 28. T-300, 301 Room ;;hermostat > g; emp. 3
- 29. Rm.114 particulate filter AP < 2.5" H 2O
. External doors
] 30o (QnSec'Nndu$h 31.. CT basin water level
. 5-10"
. }32. Automatic secondary makeup viv Auto or Open- 3
- 33. Acid day tank level Visible
- 34. CT sump pumps Operable
- 35. .P-pump (s) running
- 6. Pump strainer AP 0-7.0 psi
}37.Dischargepressure
- 8. Pump strainer AP 0-7.0 psi 39.-. Discharge pressure
- 0. . Tunnel sump puaps Operable 1.- di booster fan Running
- 2. Acid control and pH QjQOg8 / in
- 3. Blowdown control /cond. QjWg j0g8 gmin
- 44. Fission product monitor flow 95-105 cc/ min
- 5. Viv control header pressure 90-120 psig I
- 46. Pressurizer N2supply press 90-100 psig l
- 7. Check.rm. 114 from door.
8* gt f0$Nn b fi t r " red level" < 757, dark red
- 9. Seal trench {-5 n a on days.
Rev. 8/14/89 App'd h SOP /A-8b
I ; k REACTOR ROUTINE PATROL DATE: 3 50.. Full N2 bottles. Total > 0
- 51. Bank A bottle pressure > 250 psig
~
- 52. Bank B bottle pre *sure > 250 psig
- 53. Bank on service A or B I
54.- N header pressure 135-145 psi 2
- 55. Waste tank '#3 level
~
- 56. Waste tank #2 level .-
- 57. Waste tank #1 level l
- ~ ~~ '
58 Door's to Ct, WT's, Demin. Locked Rm. 114 and CT Tunnel
- 59. Time of completion of patrol
._ . 1
- 60. Operator initials
. REMARKS: . ._ -- .- _.-- - . _ . _ . - - - .
1
. . . . . . _ - . _ .n..-... . -_-_.-----
I . 1 l l l
)
I Rev. 8,L1,y@9 App'd SOP /A-8c l
I - e3 I
'g NOTE: THIS.PAGE INTENTIONALLY LEFT BLANK ii ;;ll: .I I '
l 4~ i I 3I LI _"Rg eV' < - n/1a/no App'd _ b SOP /A-8d
^5 Retyped Only
I J W PRIMARY SYSTEM NORMAL OPERATION VALVE LINEUP CHECKSHEET (cont'd)- I ,,- 56 . Valve # 568E Valve Descrig ion FE-923A valve manifold Position Intet/ outlet open;
,; 57. 595H Primary sample valve Open 58, 515U V-527A cutout valve Open l
,,59. 515AA Press drain to drain collection system. Open 1/2 turn (locked)- ]
- 60. 515B V-527C cutout valve Open -
- 61. 5155 V-5270 cutout valve Open (locked)
- 62. 544 V-545 cutout valve Open <
, 63. 599G PZR local level indicator cutout Closed (locked)
- 64. '599H PZR local level indicator cutout Closed j
- 65. 515AB Pressurizer drain to waste system. Closed
- 66. '515C P-533 suction Open (locked)
,, 67. 599A PS-938 cutout Open l
i
, ,, 68 . 5998 -PS-939 cutout Open f
, 69. 599C PS-940 cutout Open i
; , 70. 5990 PS-941 cutout Open j 71. 599E PS-945 cutout "
Open
- 72. 599F PS-946 cutout Open .
,,73. 599N FE-913A drain Closed
- 74. 5990 FE-913A drain Closed
- 75. 599V FE-913B drain Closed L- - 76. 599W FE-9138 drain Closed
~
COMMENTS: . I . Ooerator I Rev. 214j89 App'd k SOP /A-12c
MlJ y ;. ~ { i i' !
- b. ' STANDARD OPERATING PROCEDURES NANUAL f 2nd Edition, Effective Date: 5/02/89 -
Ij (Revisions.#1 through #24 to the October 1981 printing l (, > were incorporated.) ; Revision Number 2 Revision Date: January 16,1990 1 E '. Page Number ff
; SOP /I-17 .y - '
l SOP /I-21 l lg),y 3; SOP /II-14 SOP /II-15 ~ l I,? I
' SOP /Ih-11 ' ;
sop /V-4 >> f
,9 ,
SOP /V-6 .c
; 1 SOP /V-7 i
~ SOP /V-8 New-
.. SOP /VI-1
- SOP /VI-2
.? m . SOP /VI-8 'g7 > 'o' SOP /VI-9 gj SOP /Vll-29 SOP /A-1A
'I SOP /A-8a u-2 SOP /A-8b J SOP /A-14a SOP /A-14b II-3 i ' i ,' s . .
I B. Yellow Tags.
.. Yellow tags 'will' e used to' identify equipment' which, 'if operated, could presint a hazard to equipment. The tag ;
will also contain information as'to the potential: hazard. ] _ I -1.4.11.3 Equipment Tagout' I s J) ; A. Tagout Log !, -1. The tagout log will be maintained in the control room . q by the on duty shift-supervisor.
,7
- 2. The tagout log will be made up of four sections:
- a. Tagout. Instructions .)
- b. Index
- c. Active Tagout Sheets
- d. Cleared Tagout Sheets B. Performing a Tagout
.1. A tagout will be accomplished by a licensed operator I
only. L
- 2. The on duty shift supervisor nust approve the hanging
'I- or removal of any tag.
- 3. 'In the event that a tag is. missing, the shift super-l visor will be informed immediately. The missing . tag l 'will be cleared from the tagout sheet and a new tag 1 I issued. ;
l ' C. Tagout Audit
- 1. A tagout audit will .be performed at least monthly.
- 2. The tagout audit will consist of verifying that the ;
p . index includes each active tagout and also that each tag for the active tagouts is still in place and L correct. ,
- 3. IJpon completion of the audit, the person conoleting the ' audit will sign and date the tagout audit sheet.
- 4. tJpon completion of the audit, all the tagouts in the inactive section of the tagout log may be discarded.
b Rev. _1Lgj g App'd _ h ( SOP /1-17 LI
{ ^' 8
~
TABLEIVIcontinued)- - Scran Run-In Alarm Units.
- 28. Fission Pr'oduct ,.
200. cps Monitor Hi Activity. 3 see below cpay
- 29. Off-Gas Hi Activity --
- 30. Secondary Coolant -- --
10 cps-Hi Activity
- 31. Anti-Siphon Line Hi --
' c5 --
inches -] Level (above valves)
- 32. Pool Level Low >24' >28' --
feet
- 33. Rey Blade --
<10T. or <20% or >60% % withdrawn bottooed
- 34. Vent Tank Low Level --
i-11 -- inches (below C) '
- 35. Secondary Coolant -- --
<1800 gpm Low Flow
- 36. Ch 4, 5, or 6 Downscale -- --
<95 % full-scale
- 37. Valve 546 A or B -- --
off closed
- 38. Valve 509 off open -- --
- 39. Valve 547 -- --
off open
- 40. Valves 507 A/B off open --
closed with P501 on
- 41. Valve S-1 -- --
90% open or. -- 90% shut This setpoint is determined by the semiannual calibration.- Rev. 1/16/90 App'd jgypp l ,, 50P/I-21 _. _ _ _ _ _ . _ - _ _ _ _ = _ _ - ___,_________ . _ - .__ _ - _ __-
7 - . .._. i I - H. Unscrew lift cod ssembly and lif t _up as far as' it will go. 'l pg-W t Remove lift Pod and lower housing as one unit, i
- 1. Using bolt removal toel, loosen offset rr ::hanism hold down -
; b'olt on rear of assembly.
J. The lifting and removal' of the offset can be accomplished by ] ; using a center pull lifting , rod, the J-shaped "T" lif ting ] L- - tool, or in some cases both tools. If using the' center-pull: ]
'i rod, thread the rod in snug and raise the- blade.to full out. - ]
If using the "T" lif ting tool, insert the pulli_ng tool into 3
.the hole in the counter balance arm and raise the blade to -]
l full out, then insert the "T" section of the lif t ' tool' into ] the lifting lug at the top rear of the offset. ] K. Attach the lifting tool to the crane. l L. Jog the crane while lifting by hand until the offset mecha-l nism lifts free of the side guide pins. Observe clo_se_ly_ the ] strain necessary to break the offset free. If the offset .l l . mechanism does not break loose from its reflector platform- l 1 L after' applying a reasonable amount of tension, relieve the l tension on the lifting tool and determine the reason for> the-h difficulty before continuing.the attempt to lift the mecha-nism. q M. After the offset mechanism has cleared the guide pins care- , fully raise it until' the top of the blade. mount is about' 1/8" '
~
below the pressure vessel intermediate-flange. NOTE: Af ter the mechanism clears the guide pins, the blade is still partially within the gap so caution must be observed to hold the mechanism steady while
- raising it to the flange above or the blade may be L damaged.
N. With the blade now clear from its gap, carefully move'the i mechanism away from the pressure vessel " spool" flange and ig E raise the mechanism to the surface. Rev. 1/16/90 App'd g_ SOP /Il-14
u l CAUTION: The lower portion of the mechanism and the-lower tip of the blade will be very radioactive,- ; so insure close Health Physics coverage is pro-K , vided before- raising the blade to the pool. water surface. I 11.3.4 Installation of the Control Blade Offset Mechanism L The procedure for installing the blade and offset mechanism is , essentially the reverse of the above, with a particular emphasis l on the following: points: . A. Before inserting the mechanism, check the clearance of the l blade gap with the gapping tool. B. After disengaging the lifting tool, the Shift Supervisor -- ] .
. or Senior Operator will exercise the blade over its full m
length of travel until he is convinced that the blade moves j l l freely and is able to travel through the gap totally without 1 resistance. C. During the subsequent pull for the rod drop test, determine
;g 3/ the position at which the photocell for the drop timer'
. actuates with respect to the ' blade full-in ' position. It
; must not be greater than 5.2".
. 11.4 Waste Tank Analysis A waste tank analysis is performed by the laborator) +
,s for O evidence of activity prior to release to the sanito sewer.
L See Section VII.8.6. A pH of each sample is measured to i determine its acidity. If the waste water is very Nidic (pH L less than 4) and the liquid waste is to be held, a caustic 1 . solution should be 'added and circulated to prevent excess corrosion of the waste tanks. A. h I. ,. Rev. M/$ App'd _k, SOP /II-15 L
*' .r ,s, t
- 4. . Press: source; check push button and monitor point of" trip, -
verifying the following to have occurred:
- a. - Scram and rod; run-in trip actuator amplifier tripped.
b.- Building air plenum high activity scram alarm-indicated on, annunciator.
- :: c. Evacuation or isolation scram alarm indicated on-
, e
- annunciator.
- d. 16" isolation valves, indicate closed.
- e. Containa ' isolation horns'have sounded.
L f. Isolatio., soors M0-504 and H0-505 indicate closed. l g. Supply and return fans have secured.
- h. Red ficsher light outside outer containment door is flasning.
- i. Alarm burzer on ARMS module is alarming.
- 5. If more tha' one check is required, g'
B a. The horn cutout switch may be used to silence the containment horns.
- b. The 16" isolation valves cutout switch may be turned
.to the off position, leaving the vale s closed.
f c. The motor operated isolation doors may be left in the closed position.
- 6. -When the~. checks have been performed as required, reset Ii the tripped Channel 7, 8, or 9 trip.
. 7. Trip the backup door radiation monitor with the attached source. (Trip set pointer may have to be lowered to
.. obtain trip.) Verify that the items in 4 above are initiated by the monitor trip.~ Reset the monitor. ] j (Return trip set pointer to proper setpoint if moved.) 3 l
- 8. Close the 16" isolation valve cotout switch, verify the f valves indicate open.
- 9. Open isolation doors M0-504 and M0-505 by depressing the
[I, open push button for 5 seconds after the f ans start. j
- 10. Perform the visual inspection of the ventilation system on the fifth level.
- 11. Turn the ARMS detector channel selector switch to ,
Channel 5. Rev. 1/16/90 App'd ty SOP /III-11 g l
I B Remotely open valve 565B from the primary / pool drain 3. collection system control panel. Insure valve does indicate open.
- 4. The skimmer pump may be started at this point. However, I 5.
it will fill by gravity if desired. When proper pool level is obtained, secure the skimmer pump and remotely close valve 5650. Insure it does indicate closed. B. The second approved method of filling the pool is via the 4" line from tank T300/301 to the pool pump suction and discharge line. I 1. Check capacities of tanks T300 and T301 and check proper valve lineup.
- 2. With the pool system in the normal shutdown mode, fill-ing the pool through a pool pump can be avoided by opening valve V522C and permitting T301 or T300 to drain by gravity feed alone.
- 3. Close valve V522C when the filling operation is com-pleted.
V.4 Pool Lowering Procedure V.4.1 Lowe ag Pool Water Level .o Refuel Bridge I. Two methods of lowering pool level may be used: - A. By use of the skimmer system (SOP /VII.5.2), I B. By use of the pool pumps P508A or P508B as outlined below. V.4.2 Before a lowering of the pool level using P508A/B is commenced, place pool system in service as follows:- A. Isolate one pool heat exchanger utilizing the local inlet
. gate valve.
B. Place mcster switch 151 to test. 3 C. Place V509 to manual /open. I D. Start P508A or 508B. Rev. ,1L1Gfgg, App'd M , L ;, SOP /V-4
....-l L .
, B. Check valves V515M, V515X (P513B bypass) and V5150 closed for normal operation.
, NOTE: Opening V515M and closing V515T bypasses flow around valve V509, HUT-504 and P50E3/B to the input of L P5138. Opening V5150 and closing V515P. returns processed water to the suction side of P508A/B rather j than back to the top side of the pool. C. Make certain that om of three possible demineralizer units is valved into the cleanup system according to the pro-cedures described in Section Vll. i D. Turn on demineralizer flow recorder. Time and date the strip chart. E. Turn on P5138 from the control room by turning the PS13B control switch to the ON position. , F. Verify a 5015 gpm flow rate on the pool cleanup loop flow l recorder. G. Indicated flow rate should be 5015 gpm and the indicated ! water purity should be less than 2.0 unhos/cm from the demineralizer. i V.5.2 Discharging Excess Water from Primary or Pool System with T301 Full - moved to SMP-20. ' V.5 Single Pool Pump Operating Procedure ] NOTE: This procedure will be accomplished after a need has been ] determined to secure one of the pool pumps and only with ] the reactor shutdown and a radiation survey of the work ] area in room 114 completed. 3 A. Shut down the reactor. ] B. S':ure the pool system. ] C. Open thc breaker for the pump to be secured at the motor ]
. control center. ]
D. Lock out the run/stop switch at the pump to be secured. ] Rey. 1/16/90 App'd M SOP /V-6
$ W
i it V.6 - E. Close the pump suction and discharge valves (on the pump 10 ] be se:,ured). Perform appropriate system tagout. 3 1 F. Check fully open the suction and discharge valves on the ] l operating pump. ] -i G. Fully open valves 522 A/D (Pool HX discharge valves). ] H. Place the pool system back on line. ] 4
- 1. Monitor pool flow, reflector D/P pool demineralizer flow. ]
I NOTE: Trip point for reflector oP may have to be reset for single pump operations. 3
]
i I J. Adjust pool flow potentiometer. ) K. Note in the control console log that single pool pump ] i operation has commenced. ) l i l 1 I J T l-l , l I I Rev. 1/16/90 App'd h% SOP /V-7 9
.,....n..-~ - . . . ..... -. . . . . . . . . . . -.. . - . ~ . . . . -. . - . . . . ....- -.- -.._ - . . .- . . . . _ . - - . . . . . . . . - . . _ , . . ?
I . I I THIS PAGE INTENTIONALLY LEFT PLANK I . I 1 l
- - I l
l l I l i I New 1/16/90 App'd M SOP /V-8 ] i
,+,r - - - , , w %,- . - - - - - ,. . . ~ , ,-. .,...- ,,,--.%,--+,, . - -.w.-- e.. .
I I SECTION VI SECONDARY COOLING SYSTEM VI.1 Startup of the Secondary System 1 i A. Before attempting to start up the secondary system, it should be determined that:
- 1. Water level in the cooling tower basin is between 5 and ;
14 inches.
- 2. All personnel are clear of cooling tower equipment and fans. '
.. 3. Oil level in the gear reducers to the fans is normal.
- 4. The automatic sump makeup water isolation valve electrical power switch is in auto. I B. The following manually operated valves in the cooling tower should be in positions indicated: i l.
.An Closed S-17 S-9 S-105 S-129 S-18 S-10 S-106 S-101 l S-19 S-11 S-107 S-126 S-20 S-12 S-108 S-121 S-21 S-118 S-109 S-123 1 i S-22 S-119 S-110 S-125 S-155 S-120' j S-111 S-127 S-5 S-117 S-112 3-113 ] ;
S-6 S-114 S-163 S-102 ] , S-7 S-115 S-128 S-8 S-n6 I C. The following valves in equipment room 114 passageway and waste tank room should be in the positions indicated: l ,0,p,e,n Closed l S-152 S-151 S-103 S-169 S-153 S-150 S-160 ] S-104 S-159 S-170 ] Rev. 1/16/,90, App'd h _ SOP /VI-1 I .
I D. The following valves in equipment room 114 should be in the position indicated: Open Closed I S-161 S-162 S-26 S-41 S-132 S-133
~S-144 S-145 S-43 S-27 S-154 I
S-131 S-130 S '44 S-28 S-134 S-23 S-35 S-29 S-135 S-24 S-45 S-137 S-136' 5-25 S-140 S-138
- S-39 S-141 S-139 S-30 S-142 S-157 S-31 S-143 S-158 E. For operation of the chiller units with feed water from P-1, I P-2, P-3, or P-4, the following valves in room 278 should be in the positions indicated as follows:
Open Closed I S-53 S-55 S-57 S-58 S-54 S-146 S-149 ]
]
I F. Verify that the Bailey Meter recorder Ndel E101 in the reac-tor control room is on to monitor secondary flow and temp-erature during operation. Time and date chart. Secondary outlet temperature for each heat exchanger can be monitored in the control room with the digital readout and selector switch. I' G. Verify that the circuit breakers for P-1, P-2, P-3 and P-4 on MCC-2 in the cooling tower are closed and that the I. control switch on the pannels is in auto mode. I Rev. 1/16/90 App'd M SOP /VI-2 I .
i 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 I above the system setpoint, the acid injection valves automatical-ly open and acid is gravity fed into the tower sump. The acid 1 used is concentrated sulfuric acid supplied from the 250-gallon j day tank in the cooling tower. J I VI.6.4 Sample Paths for pH and Conductivity I i , A. During normal operation of the secondary system, the auto-matic pH and conductivity control units receive their sample
~
water through valves S104 and S151. '; l B. During operttion of the secondary system with the air condi- ; l tioning units secured and isolated, close S104 and open S103. This provides a representative sample for these units to control pH and conductivity. ] . l C. The pH and conductivity units shut down when secondary pumps (P-1, P-2, and P-3) are secured. l l VI.6.5 Secondary Water Corrosion Prevention 1 The prevention of corrosion in the secondary system is
~
A. accomplished by automatic addition of a corrosion inhibiter, i B. These chemicals are fed automatically by a metering pump ) I. system based on makeup flow. , J u 1 Rev. 1/16/90 App'd m , SOP /VI-8 j
e I
~
I VI.6.6 Secondary Silt, Algae and Hud 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. ! I Microbiological and algae growth is controlled by the addition of two microblocides/algaecides. The addition frequency is 3 determined by weather conditions and reactor operations.
- I.
I LI L - l 1 I- I L b 1 II I L l I ,I l
~~
- Re v . M1,6/,9,0, App ' d _ ,,
50P/VI-9
tv- _c L Vll.8.12 Chemical Precipitate Treatment A. Drain the waste tank to WT2. B. Lower the waste tank water pH.
- 1. Check all valves at R200 closed. l
- 2. Line up acid mixing tank valves and close the pump breaker. 4 I 3.
4. Open R200 valves RE57, P.E58, RES and RE70. Start acid pump at R200 station and throttle flow with l RE58.
- 5. Add sufficient acid (6 normal) to lower pH to between i 5.0 and 6.0.
- 6. Secure the acid pump and opn the breaker.
- 7. Shut valves RES, RE57, RE58 ai.d RE70.
- 8. Drain and flush the acid mixing tank.
I. ^*
- 9. Close the acid mixing tank valves.
C. Sparge and recirculate, bypassing the filters, for 30 : minutes. r D. Add a special carrier solution which will Le provided by the ' Laboratory Group. E. Sparge and recirculate, bypassing the filters, for one hour. F. Reise the pH.
- 1. Open the UT2 manhole cover.
g_ 2. Add sufficient sodium hydroxide to raise *.ne pH to , W 11.0-14.0. ,
* *. CAUTION: It is better to add too much than not e'nough.
- 3. Replace the manhole cover. -
G. Sparge and recirculate, bypassing the filter, for 30 minutes. H. Secure W.T. recirculation and let tank settle for 24 to 48 i I hours. Then pump (without air sparge) through stand pipe, fron WT2 to WT1. I I. WT1 should now be ready to sample. J. Recirculate WT2 through filter until they no longer foul up, ]
- disposing of used filters in drying rack. ] .
K. Remaining water in W.T. #2 should now be ready to sanple. ] Rev . JL13/Sp,, , App ' d _%, SOP /Vi!-29
I REACTOR STARTUP CHECKSHEET FULL P0blER OPERATION BUILDING AND MECHANICAL EOUIPMENT' CHECKLIST DATE: TIME (Started): l I 1. Emergency air compressor (load test for 30 minutes after maintenance day).
- 2. Beamport Floor:
II a. Beamport radiation shielding (as required).
- b. Beamport status checked / updated. ] !
I 3. c. d. a. Seal trench low level alarm tested (after maintenance day). Check closed beamDort floor access gates. Check operation of ion failure buzzer and warning light. (Required if
]
I 4. shutdown longer +han 4 hours.)
- b. Test stack ms %or and low flow alarm per 50P while in west tower.
Emergency generate. a', ailability/ status checked. (If shutdown for greater ] 1 l i than 24 hours, run er.ergency generator for 30 minutes.) ] I 1 5. Emergency pool fill. (Check valves Ply-1 and P!V-2 locked open.) i
- 6. equipment: i Visual
- a. Oil check level inofCTCTfans andnormal secondary (af ter maintenance day). i I 7.
- b. Secondary makeup isolation valve power switch closed, verify operation and placed in auto mode.
Visual check of room 114 equipment:
~~
valve cycled to i
)
l
- a. P501A and P5018 coolant water valves open.
- b. Pump controllers unlocked to start (as required).
- c. Check valves 599A and 599B open. l
- d. Air valve for valve operating header (V0P31) open. I
.I e. f. N2 back-up valve open. Ai r/N2 cross connect valve open. S1 and 52 hydraulic pumps on (oil level normal). l
)
- g. .
- h. Valves S1 and S2 cycled in manual mode and positioned as required. l
- 1. Vent the pool hold-up tank. l
- j. Vent the pool skimmer system pump. I
~
- k. Check the pipe trench free of water--check the four-pipe annulus drain ] I valves for water leakage after maintenance days. ]
- 1. Add D1 water to beamport and pool overflow loop seals,
- n. Check oil reservoir for pumps 501A, 501B, and 533 for adequate supply.
Add if necessary. II I 8.
- m. Visually check room 114 and DI area after all systems are S. operation.
Reactor Pool-
- a. Reflector experimental loadings verified and secured for start-up.
I b. Flux trap experimental loading verified and secured for start-up, or strainer in place. Check power on and reset, es necessary, silicon 4tegrator, totalizer I I c. I REACTOR CONTROL SYSTEM CHECKLIST 1. setting, silicon rotatar and alav system. All chart orives on; charts timed and dated. IRM recorder to slow. Fan f ailure warning system cleared. I 2. 3. 4. Annunciator board energized; born off. Television receiver on.
- 5. ' Primary / pool drain collection system in service per SOP. (Manually pump DCT)
- 6. Secondary system on line per 50P (as needed).
~ ^
- 7. Primary system on line per SOP:
I I a. Primary cleanup system on line. I 8. Pool system on line per SOP:
- a. Pool cleanup system on line.
- b. Pool reflector aP trips set as required.
- 9. Nuclear Instrumentation check completed per SOP:
l I a. The following trip values were obtained during the check: IRM-2, run-in seconds (11+1) Scram seconds (9+1) 1RM-3, run-in ~~~" seconds (1171) ~~ Scran ~'~~~" seconds (971) l WRM-4, run-in ~ ~ % (114+1) Sc rar. ~~"" % (119+1) ~ W PRM-5, run-in ~ ~ ~~ % (11471) Sc ran ~ ~~" 1 (11971) PRM-6, run-in ['"~ % (11431) Scran ,t(11911) Rev. 1/16/90 App'd h _ , SOP /A-la
I. REACTOR ROUTINE PATROL Date: _ ,,,,,_ _ ,,,, m ' I. Time of start of patrol J
- 2. Time and date all charts I, 3. Check ARMS trip settings
=- --- .--- -. _ .._.___..... --- -
I 4. 5. Visual check of entire pool Anti-siphon tank pressure 36 psig i 3 psi I. 6. North iso door seal pressure 18-28 psig 7.. South iso door seal pressure 18-28 psig
~~ ~
- 8. 5th level backup dcors Open I _9 . .-
St3 level detector reading 10, 5th level trip point set 0-3.5 mr/hr 3.5 mr/hr 4
- 11. 16" iso viv A air pressure 45-35 psig
- 12. 16' iso viv B air pressure > 90 psig
- 13. Emerg compress on standby ,c]sg ,
g open,
' ~
- 14. Containment hot sump pumps Operable
- 15. Door 101 seal pressure 18-28 psig.
~
- 16. BP ficor Conditions normal.
- 17. Fuel vault Locked g ____ ____ ____ _ _ _ _ _ _ _ _ _ _ _ _
E 18. Inner airlock door seal press. 18-28 psig
- 19. Outer airlock door seal press. 18-28 psig I 20. Cold deck temperature 4 5' + 55' E,.-.
-3
- 21. T~300 level > 2000 gal.
- 22. T-301 level < 6000 gal.
~
- 23. Labyrinth sump Level < Alarm Pt. I
= _ - - - . .
- 24. RO unit power ON Uii~tTETirst routTne patroT ofInT'da7'or~tTe T1rst patroT'aTt'iFT'sTartup,drTin ~aif ~ ~
water from the anti-siphon system. If draining causes the pressure to drop signifi-cantly, return to the middle of the band (36 psig) and record the pressure here, if a condition or reading is normal, enter a "'" (for conditions) or the reading in the applicable box. If the condition is abnormal, enter the condition or reading and circle it. Explain all abnormal conditions or readinge, in the REMARKS on page 3. I Rev.jfjpf 9p, App'd h, 50P/A-Ba I
REACTOR ROUTINE PATROL Date:
- 25. RO unit temperature 24-28'C or Standby
- 26. RO unit pressure 190-200 psig or i Standby ;
27a. UPS RM No Alarms Indicated ] Thermostat > 55'F Temp > 40'F 27b. T-300, 301 Room jne mostat > g !
] .
Perf orm complete Thermostat > 60"F " I zue. LU Nm. 28b. Battery charging current (checklist--SUNDAY) Temp
< 1 amp.
> 50'F ,'
]
Z8c. Battery Voltage > 28V
~
]
- 29. Rm. 114 particulate filter & < 2.5" H 2O
'S !W
- 30. External doors All locked except east when sec on duty.
l l 31. CT basin water level 5-10" W 32. Automatic secondary makeup viv Auto or Open l
- 33. Acid day tank level Visible ,
l i
- 34. CT sump pumps Operable j
- 35. P-pump (s) running
- 36. Pump-strainer # 0-7.0 psi
- 37. Discharge pressure
- 38. Pump strainer & 0-7.0 psi I .-.
- 39. Discharge pressure I l
EA
- 40. Tunnel sump pumps Operable i l
- 41. WT booster f an Running l
- 42. Acid control and pH Flow 400-800 cc/ min t- (Range as posted)
- 43. Blowdown control /cond. gy0g8g/ in
- 44. Fission product monitor flow 95-105 cc/ min ;
l 45. Viv control header pressure 90-120 psig
- 46. Pressurizer N2 supply press 90-100 psig
- 47. Check rm. 114 from door.
46 Deltecn oil filter 'rea level' l< 75% dark red I ! ! ! I and blowdown j l j ; I l o I49. Seal trench 61-66' Run pump on days. i i i
- l. 1 l l 1/16/90 isp p ' a Uwm 50P/A-So IiFev.
,,i, , , . ,, _ - . - - , _ . .
l t Date POOL SYSTEM VALVE LINEUP CHECKSHEET I This chae.ksheet shall be completed when required by the SOP. The operator performing the chews will verify the position of each valve and indicate the verification by initialling the checksheet. Under the direction of the Shift Supervisor, a valve may i i be positioned other than noted on this sheet. The, operator will check the valve to be in the desired position, line out the normal position on this sheet, and write in the l actual position _of the valve. The reason for the valve being positioned abnormally will be noted in the COMMENTS section. .
. i Throttled valves shall be checked to be in the position shown by the tag on the valve. .
Note the valve's position in the space provided on the checksheet. l Valve A Valve Descriplion Position
- 1. 598A Air supply to 546 Open i
_,_2. 555P Air supply valve to V-547 Closed
- 3. 548A P-532 pool suction Opea i
- 4. 548B P-532 pool (at refuel) suction Closed
- 5. 518U Vent valve (pool TH ) Closed ]
I _,_ 6. 7. 515X 515N P-513B bypass P-513B discharge Closed Open
]
l
- 8. 518H P-513B suction gage catout Open l
- 9. 518G P-513B discharge gage cutout Open
_,,10. 522C Pool drain / fill Closed ) l _ ,, 11. 522B Pool fill Closed ] Cleanup return to loop _ ,,12. 51s' Q Closed )
' l
__,13. 515M Cleanup suction from pool Closed ) l _ _ 14. SIST Cleanup suction from loop Open _ _ 15. 522F P-508A discharge Open __ 16 . 522E P-508B discharge Open
- 17. 5318 P-508B bypass Closed
- 18. 5181 P-508A gage cutout Open .
_ ,, 19. 518J P-508A gage cutout Open _,,20, 518AD P-508B gage cutout Open _, , 21. 518AC P-508B gage cutout Open j _ ,, 22. 599J PS-947 cutout Open ] l _ ,_ 23. 539A P-508A suction Open ] _ ,, , 2 4 . 539C P-508B suction Open ] l Rev. yjypp, App'd gg, 50P/A-14a l L I l
<I<:
4
; POOL SYSTEM VALVE LINEUP CHECKSHEET.(cont'd)~
Valve # Valve Description Position
-25. 518V Vent valve Closed ~)
- 26. Cleanup return to pool I
515P Open
- 27. .514B HUT outlet Open
- 28. 539B HX-521A inlet Open
- 29. 539D HX-521B inlet Open
- 30. . 522A HX-521A outlet Throttled )
- 31. 522D HX-521B outlet Throttled 3
- 32. 518N HUT vent cutout Open
- 33. 518K HUT vent Closed J.
[ 34. 515R HUT drain Closed (locked)
- 35. -5150 HX-521A drain Closed )
- 36. 515Z HX-521B drain Closed -)
- 37. 518W Vent valve Closed )
- 38. 518AG Pool system "Y" strainer drain Closed
- 39. 518Q Drain valve-(tunnel) Closed )
40.- 518R Drain valve (tunnel) Closed )
._ 41. 514A V-509 cutout Open ]
- 42. 568G PT-917 cutout valve Open
=
- 43. 599Z PT-917 vent Closed 44, 599Y P-532 suction vent Auto Float 45, 5151 P-532 suction Open L 46. 5180 P-532 gage cutout Open l 47.
48.
.518C 515E P-532 gage cutout Skimmer filter inlet Open Open 1 I-49.
50. 5150 567A
' Skimmer filter outlet Drain collection pump suction Open Open I ,g 51. 567B Drain collection pump suction drain Closed (5; 52. 566 Drain collection system discharge Open
- 53. 567C Drain coll sys disch to 513B suction Opr ;
hl 54. 593 Skimmer suction T-300/T-301 Open
- 55. 518AJ/AK HX-521 A/B loop vents Closed )
I
-l Rev. 1/16/90 App'd \p S0P/A-14b l
i
a . .__ I 2-) ,
, MURR STIE EMERGENCY PROCEDURES AND FACllIIY EMERGENCY PROCEDURES Revision Number 3 Revision Date: September 18,1989 Section Page I- HugLhtt Number SEP 2 2 SEP-2 3 SEP-3 3 SEP-4 3 SEP-7 1 l: SW-7 2 .
SEP-7 3
.I; SEP-7 4 EMERGENCY CALL LIST FEP-3 1 I-I g
I I I. I 11- 4 I . l
. I I SEP-2 (Cont'd) Page 2 of 5 j
- 5. Send operator to west tower with radiation monitor to:
NOTE: Communicate by intercom, sive stack nonitor is affected by 1 portable radio RF. l
- a. Verify radiation background at stack monitor. l i
- b. Verify control room readings.
'I c. Mark initial needle positions on ani.'og display with the time for future
.W analysis if the control room display becomes inaccessible.
i
- d. Verify flow rate through monitor is 7,+ 1 SCFM.
If not, use Worksheet A to determine stack monitor values. l l L 6, If nuclides which are being released are in doubt, pull stack filters and analyze. , I L 7. The EMERGENCY COORDINATOR shall evaluate the need to evacuate specific portions of the facility. rl-
- 8. The EMERGENCY COORDINATOR shall appoint and have a surveillance team check any areas evacuated in step 7, clear of personnel within 30 minutes.
! NOTE: EMERGENCY DIRECTOR approval required for any voluntary radiation c exposure in excess of 10CFR20 limits. (Vp to 100 rem for life-h saving, up to 25 rem to prevent exposure to members of general public in excess of 1 rem whole body and 5 rem thyroid.)
- 9. The EMERGENCY DIRECTOR shall determine the need for EMERGENCY SUPPORT m- ORGANIZATIONS and. if needed, activate them or place them on standby.
See Table 11. EMERGENCY S',30RT ORGANIZATIONS. E l L NOTE: If 9-911 is cilled during any emergo.cy, contact MU News Bureau. ] I l Rev. 9/18/89 App'd M Y
Page 3 of 5 I SEP-2 (Cont'd).- TABLE II j EMERGENCY SUPPORT ORGANIZATIONS
- a. UMC HEALTH PHYSICS SERVICES In the event of a miiological emergency, the UMC Health Physics i Services may be contacted te assist in enecking facility personnel for ,
contamination. After hours the Watch Office may be contacted to open i the Research Park Development Building (backup emergency control center). One of the persons, listed below will man the backup control center. ,q CONTACT Office Home Dr. Philip Lee 882-7221 445-5275 ' 4 Jamison Shotts 882-7221 474-2194 David Spate 882-7221 657-9450
- b. UMC POLICE i The UMC Police may be called to restrict entry 24 hours to the facility. 882-7201
- c. UNIVERSITY OF MISSOUR,l_ HOSPITAL AG Ambulance CLINIC 5 (UMH&C) 682-6003 E- The UMH&C should be contacted in the event Walk-in- .)
9 of personal injury. In the event of personal 882-8091 . contamination or radiation exposure without ; 'e iyury, see MEDICAL. EMERGENCY PROCEDURES. q ,E If three or more personnel are involved, ask the , Administrator-0n-Duty to implement the Radiation i I' Disaster Plan. Refer to the MEDICAL EMERGENCY PROCEDURES for details. 3 d. MU NEWS BUREAU See SEP-7, ] g- PUBLIC . This office will initially deal wi+,n questions INFORMATION from offsite. Direct any questions from media to PROCEDURE I this office. They will release sta+,ements only by EMERGENCY DIRECTOR authorization, ' I e. COLUMBIA FIRE DEPARTMENT The Columbia Fire Department shall be notified 24 hours WT~ in the event of fire or need of emergency rescue capability. Insure Office of University Relations is also called. Rev. 9/18/89 App'd g_ I m_.um ---__m -___.__a _m 4 m m 2 - .- oa-m
=. . . .
I l ' SEP-3 (Cont'd) Page 3 of 6 TABLE 111 EMERGENCY SUPPORT ORGANIZATIONS
- a. UMC HEALTH PHYSICS SERVICES In the event of an ALERT condition, the UMC Health Physics Services may be contacted to man the backup emergency control center.
After hours, call the Watch Office to open RPDB. CONTACT _ . Office Home . I Dr. Philip Lee Jamison Shotts David Spate 882-7221 882-7221 882-7221 445-5275 474-2194 657-9450
- b. UMC POLICE The UMC Police may be called to restrict entry 24 hours to the research park and to assist in partial site 882-7201 I:' area evacuation if deemed necessary.
- c. UNIVERSITY OF MISSOURI
~ ~~
HOSPITAL AND Ambulance CLINICS (UMH&C) 882-6003 The UMH&C should be contacted in the event Walk-in of personal injury. In the event of personal 882-8091
- contaminatir,n or radiation exposure without injury, see MEDICAL EMERGENCY PROCEDURES.
If three or more personnel are involved, ask the I Administrator-On-l'uty to implement the Radiation Disaster Plan. Refer to the MEDICAL EMERGENCY PROCEDURES for details. I d. MU NEWS BUREAU This office will initially deal with questions See SEP-7, PUBLIC INFORMATION
)
from offsite. Direct any questions from media to PROCEDURE this office. They will release statements only by EMERGENCY DIRECTOR authorization.
- e. COL _UMBIA FIRE DEPARTMENT 24 hours 9-911 The Columbia Fire Department shall be notified in the event of fire or need of emergency rescue capability. Insure Office of University Relations is also called.
I Rev. 9/18/89 App'd g I I .
, _ . . . . - -... i
.,m.
.J ;
j SEP-4 (Cont'd) Page 3 of 6 i i TABLE IV l EMERGENCY SUPPORT OfiGANIZAT10NS 1
- a. UMC HEALTH PHYSICS SERVICES I in the event nf a SITE AREA EMERGENCY, the UMC Health. Physics Services may be contacted to man the backup emergency control center, i
j After hours, call tne Watch Office to open RPDB. CONTAC_T Office Home j Dr. Philip Lee 882-7221 445-5275 Jamison Shotts 882-7221 474-2194 i David Spate 882-7221 657-9450 1
- b. UMC POLICE The UMC Police may be called to restrict entry I to the research park and to assist in partial site area evacuation if deemed necessary.
24 hours _ 882-7201 1 i
- c. UNIVERSITY OF MISSOURI ~~~
HOSPITAL AND Ambulance I ELINIC5(UMH&C) 882-6007" ] The UMH&C should be contacted in the event ' Walk-in of personal injury. In the event of personal 882-8091 contamination or radiation exposure without , injury, see MEDJCAL EMERGENCY PROCEDURES. If three or more rprsonnel are involved, ask the > I Administrator-0n-Duty to implement the Radiation Disaster Plan. Refe? to the MEDICAL EMERGENCY PROCEDURES for deta'is. i l I d. MU NEWS BUREAU This office will initially deal with questions See SEP-7, PUBLIC INFORMATION 3 from offsite. Direct any questions from media to PROCEDURE this office. They will release statements only by EMERGENCY DIRECTOR authorization.
- e. _ COL _UMBIA FIRE DEPARTMENT. 24 h_ours 9-911 The Columbia Fire Department shall be notified I in the event of fire or need of emergency rescue capability. Insure Office of University Relations is also called.
I Rev. 9/18/89 App'd M, I
Page 1 of 4 I SEP-7 i PUB _LIC INFORHATION PROCEDURE NOTE: The MU News Bureau shall be activated to handle the release of public ] L3 information as required in the ALERT or SITE AREA EMERGENCY procedures; l3 whenever offsite emergency assistance is requested via 911; or whenever ! deemed appropriate by the EMERGENCY DIRECTOR. A. INITIAL RELEASE OF PUBLIC INFORMATION , iI 1. The Emergency Status Report shall be completed and approved by the EMERGENCY DIRECTOR. l's 2a. During normal University office hours, activate the MU News Bureau by ] Lg calling 882-6211, 882-6214 (Mary Still) or 882-9142 (Marty Oetting). ] l 2b. At other times, call the following list of MU News Bureau staff in order ] I until one of the individuals listed is reached (E their spouse, children, ) etc.).
- 1) Mary Still (314 -875-4730 ]
l
. 2) Marty Oetting (314 -474-5126 3
- 3) Ken Brogdon (314 -442-5260 ]
(4) Helen Fiengo (314 -442-8046 ] 5
- 3. Read the Emergency Status Report as appreved by the EMERGENCY DIRECTOR l j to the MU News Bureau staff member and :.nswer any questions concerning ]
definitions, terms, units. e,tc. 1
- 4. Record other questions that the MU News Bureau staff member may have. ]
Enter the name of the MU News Bureau staff member contacted and give the J completed report to the EMERGENCY C0ORDINATOR to be kept with the records of the EMERGENCY. 1' l I LI l 1 I Rev. 9/18/89 App'd l)WhT\ l l
i m SEP-7 (Cont'd) Page 2 of 4~ aI 5. The PLl News Bureau staff member contacted should verify a call concerning ] an amergency at the University of Missouri Research Reactor by calling
'I 882-4211 or 874-4119 and ask to speak to a member of the Facility Emergency Organization (FEO). If the person answering the phone does not know who is in the FEO, then ask for anyone from the Director's Office, Operations Health Physics, or Reactor Chemistry groups. The I individuals in these groups are listed below in alphabetical order.
After verifying the person's identity by asking for his social security number, the emergency call can be verified. VERIFICATION LIST FOR MJRR EMERGENCIES Name Soc. Sec. No. Name Soc. Sec. No.
~
Chuck Anderson Sue Langhorst Joe Baskett Charlie McKibben r' Rita Bonney Walt Meyer Kenneth Beamer Steve Morris Barry Bezenek Phil Neel ) I Ron Dobey Chester Edwards John Ernst Leslie Powell Mike Randolph Bill Reilly l Christine Errante Tony Schoone Mac' Evans Jim Schuh Les Foyto Tom Seeger John Fruits Vickie Spate Greg Gunn Ray Stevens I- Robert Hudson Nolan Tritschler l Rolly Hultsch Robert Walker ] y Brenda Johnson Mike Wallis Vernon Jones Tim Warner Mike Kilfoil Burle Warren -) q Ron Kitch
- 8. MU News Bureau personnel contacted will determine the need for staffing and )
equipping an emergency information center and will call in the required I. staff ano arrange for necessary facilities.
]
l
- 7. MU News BJreau personnel will inform news media and others of the public, ]
I 8. as necessary, of the emergency. If possible, a MU News Bureau staff member will be sent on site to assist ] the EMERGENCY DIRECTOR with the release of information.
?
P LI Rev. 9/18/89 App'd M L
~ - , , - - - - + - -
SEP-7 (Cont'd) Page 3 of 4 ) B. SUBSEQUENT RELEASE OF PUBLIC INFORMATION The nature of the emergency and the required response (fast or slow moving) may affect the release of subsequent information. . ' (
- 1. If time per: nits, fill out an EMERGENCY STATUS REPORT for each release 1 of information.
- 2. If time does not permit filling out an EMERGENCY STATUS REPORT for each release of information, the EMERGENCY DIRECTOR may verbally approve s i
E information to be released by the ifJ News Bureau staf f personnel on site. ] , 1l 3. MU News Bureau will provide updates on information concerning the ] ; emergency to the general public and media at periodic intervals or as it ' becomes available from the EMERGENCY DIRECTOR.
- 4. In the event of injury to personnel, the EMERGENCY 0! RECTOR will be responsible for contacting relatives. "I News Bureau staff will not ]
I 2 release names of injured personnel unti. cleared to do so by the EMERGENCY DIRECTOR and only after relatives or next of kin have been notified.
- 5. HU. News Bureau will continue its operations until the emergency is ] I
! officially terminated by the EMERGENCY 0! RECTOR or until emergency information services are deemed to be no longer necessary.
- 6. Following the terminatior of the emergency,ifJ News ;reau will arrange a ]
meeting between emergency officials and the news media. MU News Bureau , ) l will also conduct a critique of its activities and will seek feedback 1 l from the media and public on the effectiveness of its procedures. 1 C. RECORDS AND EMERGENCY LIST VERIFICATIONS
- 1. All EMERGENCY STATUS REPORTS shall be maintained by the EMERGENCY DIRECTOR or EMERGENCY C0ORDINATOR as a permanent record of the emergency.
l .
- 2. The MU News Bureau staf f call list and the MURR verification list will be ]
l reviewed and revised annually to keep the names and numbers current. 1 I ll I. Rev. 9/18/89 App'd M ,,,
.w- . , - . . . - .
k? y
; SEP-? (Cont'd) Page 4 of 4- ;
. EMERGE _N_C_Y STATUS REPORT Date: , , _ , ,
MV NEWS BUREAU STAFF MEMBER CONTACTED: ) THIS (IS11S NOT) A ORILL A. DESCRl_PTION OF EMERGENCY l
- 1. What happened and where specifically did it happen? i
,g (i.e. reactor containment or laboratory building)
W !
- 2. When did it happen?
- 3. Why/how did it happen?
lI 4. Releases, if any, of radioactive material onsite or offsite? i 1 1 j J B. EMERGENCY ASSESSMENT
~
l 1. Exten't"of~ damages and/or injuries? q
- 2. Extent of external danger to general public?
lI
- 3. Actions taken to protect the general public?
l 4. Emergency classification declared? 1 l C. ADDITIONAL INFORMATION l
- 1. Wrien will more details be available?
i
- 2. Wnen can media speak with EMERGENCY DIRECTOR?
E W i Filled in by:
]
l Approved by: !
UfGTGENCY DIRECTOR ]
I Time Date 1
~~
Rev. 9/18/89 App'd \)4m. I l 1
, _. I
M- M M M M M M~E - M M W W W W W W W W EMERGENCY PROCEDURE EMERGENCY CALL LIST _ ,,,_Ofr ons
,e,rall o, Emergency Support Organfzat_1o_ns
. . . 1!e.a,1,t,h, ,P,hy,s,1,c,s, , , , , , , , ,, - _,__
Phone No. Phone No. Phone No. ll2?lI6ff M. Evans 698-2450 UMC Police 882-7201. S. Langhorst R. Stevens 442-2539 K. Beamer 682-5499 ] J. Ernst 445-5621 ] R. Kitch 696-3710 ] Columbia Fire Department 9-911 R. Dobey 443-4513 3 C. Kribbs 682-3930 J. DeMers 445-2204 UM Hospitsi and Clinics T. Seeger 875-8656 Ambulance 682-6003 J. Baskett 474-2046 ] or 9-911 Director's Office 882-8091 Phone No. Walk-in S. Morris TE-Tfl7 - J. C. McKibben 442-6728 UMC Health Physics (Office) 882- 7221 Dr. Phil Lee (Home) 4'a-5275 Jamison Shotts (Home) 474-2194 Re>~. tor Chemisiry, David Spate (Home) 657-9450 _ ,, , , 0p,er,a,t i on s Phone No. Phone No. W. A. Meyer iPIf?T676 M. Glascock 474-8390 MU News Bureau 882-6211 ] C. Edwards 443-7529 J. Schuh 874-3086 See Public Information 882-6214 ] T. Schoone 474-6416 V. Spate 657-9450 Procedure for other or R. Hultsch 442-6653 phone numbers. (SEP-7) 882-9142 ] C. Anderson 696-5506 State Emergency Management B. Bezenek 445-5680 Agency _(SEMA) 314-751-2748 G. Gunn 875-1162 474-9388 NRC, Region III 312-790-5500 N. Tritschler L. Foyto 446-0491 (ANI) 203-677-7305 J. Fruits 474-0774 R. Hudson 875-0451 V. Jones 445-2543
- n. Kilfoil 474-6285 P. Neel 442-8693 ]
M. Randolph 442-5315 ] R. Walker 445-8077 ] M. Wallis 443-8764 T. Warner 816-882-6740 B. Warren 445-2204 ] Rev. 9/18/89 App'd W __ i
Page 1 of 1 FEP-3
]
FIRE PROCEDURE I 1. Any individual discovering fire shall notify reactor control (#13) of fire, giving nature and location of fire. The Shift Supervisor will activate the FACILITY EMERGENCY ORGANIZATION by page system and provide warning to stay i clear of fire location.
- 2. SHIFT SUPERVISOR will call (9-911) to notify Columbia Fire Department.
, 3. EMERGENCY DIRECTOR will investigate the fire and determine steps to minimize hazard to both personnel and property.
NOTE: An assessment of offsite radiological consequences shall be ; determi ned.- This assessment may require escalating emergency response to a site emergency procedure (Unusual Event, Alert).
- 4. The EMERGENCY DIRECTOR may contact the MU News Bureau to handle public 3 information, if appropriate.
- 5. If the fire cannot be put out immediately with local fire extinguishers - the I reactor WILL be shutdown to focus on fire.
- t
- 6. Secure EF-13 and EF-14. i
- 7. Secure ventilation supply and exhaust fans and close all fire doors.
- 8. If the fire is in containment and cannot be immediately brought under control, initiate reactor isolation.
I I I I Rev 9/18/89 App'd M i L lI L .
I I MURR STIE EMERGENCY PROCEDURES AND l FACIIIIY EMERGENCY PROCEDURES Revision Number 4 g Revision Date: December 8,1989 Section I Number Page Number SEP-2 1 , SEP-3 1
'b g SEP-4 1 SEP-7 2 i EMERGENCY CALL LIST
- g. , WORKSHEET C 1
I ' I ) I l L I ' I I I 11 5 f
I. ! Page 1 of 5 SEP-2 i UNUSUAL _ EVENT PROCED_URE i L ACTIONS _ _L_EV_EL_S : , \ L IF 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 radioactitity at the stack monitor exceed- ,
ing 3800 MPC when averaged over 24 hours. (See REACTOR EMERGENCY -] 1 j PROCEDURE (REP-21).] . NOTE: USE OVERLAY I TO DETERMINE EXT:NT OF ACTIVITY FOR l' L ' IODINE AND PARTICULATE. USE OVERLAf !! TO DETERMINE l EXTENT OF GASEOUS ACTIVITY.
- d. The projected concentration of airborne radiological effluents at the ;
l distance corresponding to the nearest site boundary exceeding 10 MPC when averaged over 24. hours. ! e. Prolonged fire cr explosion within the facility that can result in a release of radioactivity that would cause exposures of the publi.c or staf f dpproaching 1 rem whole body or 5' rem thyroid. l 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. L THEN,at least an UNUSUAt. EVENT condition exists. . IMM_ED_I ATE AC'.11_0_NS:
- 1. Activate the Facility Emergency Organization (FEO), as per ACTIVATION OF -
FA_CIL ITY EMER_G_EN,C,Y,,0,R,G,A,NI_ZATl_0_N_ _PR0_C_EDURE_ if not already activated. I 2. Operations shall provide information to the EMERGENCY DIRECTOR / EMERGENCY C0ORDINATOR.
- 3. If airborne activity is involved, continue with step 4. I,ff, no,t,, go to step 7.
- 4. Time and date stack monitor charts for reference.
I Rev. 12/08/89 App'd j % , I !I l-l
E. l Page 1 of 6 S_E Py3_ ; ALERT PROCEDURE ACTIONS LEVELS: i
. E there is: !
a) Concentration of airborne radioactivity at the stack monitor I exceeding 19,000 MPC when averaged over 24 hours. [See REACTOR EMERGENCY PROCEDURE. (REP-21).) 3
]
I NOTE: USE OVERLAY I TO DETERMINE EXTENT OF ACTIVITY EDR IODINE AND PARTICULATE. USE OVERLAY 11 TO DETERMINE EXTENT OF GASEOUS ACTIVITY. ; b) The projected concentration of airborne radiological effluents at the distance corresponding to the nearest site boundary exceeding 50 MPC when averaged over 24 hours. I . c) Radiation levels at the distance corresponding to the nearest site boundary of 20 mrem /hr for 1 hour whole body or 100 mrem f thyroid dose. - d) Loss of physical control of the facility. e)r Other plant conditions exist with a level of significance' of a - major failure of fuM cladding but primary and containment boundaries exist t we releases.
- THEN at least an ALERT condition exists.
~
IMMEDl_A_TE AC_TIO_NS_:
- 1. Activate the Facility Emergency Organization, as per ACTIVATION OF FACILITY EMERGENCY ORGANIZATION PR0rEDURE, if not already activated.
I 2. Operations shall provide information to the EMERGENCY DIRECTOR / EMERGENCY C00RDINATCR.
~
_3. Time and date stack monitor charts for reference.
- 4. Shut down the reactor.
Rev.12/08/89 App'd _ M Q ______ t
I Page 1 of 6 KP,-4 I SITE _ AR_E_A EMERGEN_CY Po CED_URE _ ACT_I_0_N_S_ LE_VELS : E there is: a) Concentration of airborne radioactivity at the stark monitor exceed-ing 95,000 MPC when averaged over 24 hours. [See REACTOR EMERGENCY ] PROCEDURE (REP-21).] ] NOTE: USE OVERLAY I TO DETERMINE EXTENT OF AC' '/ITY FOR IODINE AND PARTICULATE. USE OVERLAY 11 TO DETERMINE EXTENT OF GASEOUS ACTIVITY. b) The projected concentration of airborne radiological ef fluents at the I distance corresponding to the nearest site boundary exceeding 250 MPC when averaged over 24 hours. I c) Radiation levels at the distance curresponding to the nearest site boundary of 100 mrem /hr for 1 hour whole body or 500 mrem thyroid dose. d) Other plant conditions exist with a level of significance of major f uel damage ano conditions that indicate actual or imminent f ailure of containment integrity ond primary system integrity. THE_N a SITE ARE A EMERGENCY condition exists. IMMED_1 A_TE ACl10NS:
- 1. Activate the Facility Emergency Organization, as per ALTIVATION OF FACILITV EMERGENCY ORGANIZATION PROCEDURE, if not already activated.
I 2. Operations shall provide information to the EMERGENCY DIRECTOR / EMERGENCY COORDINATOR.
- 3. Time and date stack monitor charts for ref erence.
- 4. Shut vn C.e reactor.
I Rev.12/08/89 APP' d M. s g E __U
.,I SEP-7 (Cont'd) Page 2 of 4
'g 5. The MU News Bureau staff member contacted should verify a call concerning g an emergency at the University of Missouri Research Roactor by calling 8B2-4211 or 874-4119 and ask to speak to a member of the Facility
- Emergency Organization (FEO). If the person answering the phone does E not know who is in the FEO, then ask for anyone from the Director's E Office, Operations, Health Physics, or Reactor Chemistry groups. The individuals in these groups are listed below in alphabetical order. a After verifying the person's identity by asking for his social security 2 g number, the emergency call can be verified. VERIFICATION LIST F.0R MURR EMERGENCIES Coc. Sec. No. Name Soc . Sec . No. Name [ g Chuck Anderson Sue Langhorst b g Joe Baskett Charlie McKibben f Rita Bonney Walt Meyer Kenneth Beamer Steve Morris
- Barry Bezenek Phil Nee)
Joe DeMers ] Leslie Powell y Chester Edwards Mike Randolph I John Ernst Cnristine Errante Nac Evans Bill Reilly Tony Schoone Jim Schuh h a Les Foyto Tom Seeger g John Fruits Vickie Spate Ray Stevens
- Greg Gunn Robert Hudson Nolan Tritschler
- Rolly Hultsch Robert Walker Brenda Johnson Mike Wallis Vernon Joaes Tim Warner
. Mike Kilf oil Eurle Warren
.on Kitch I 5. ft' News Bureau personnel contacted will determine the need for st3ffing and equipping an emergency inf ormation center and will call in the required E staff and arrangt for necessary facilities.
MU News Bureau personnel will inf orm news media and otners of tre public, 7 as necessary , of tne emergency.
- 8. It possible, a MU News Bureau staff member will be sent on site to assist the EMERGENCY DIRECTOR witn the release of information.
s b Rev. 12/08/89 Apr. ' d %_
?I r _, .,, n +n x. a c .
u,
l .W' M' W W W i M - -_M ' M' W W W W! ' W. m , W :L W . W
- p. .
EMERGENCY PROCEDURE EMERGENCY CALL LIST
. 2. . . .H,e,a,1,t,h, Phy,s,i c s , 0pe,r,a,tions - ~ Emergency Supp, ort _ Organization g Phone flo,. Phone No. Phone l _.-'
S. Langhorst T47'ISY4' N. Evans 698-24^66 UMC Police 882-7201 R. Stevens 442-2539 ~K.'Beamer 682-5499
^~~
J. Ernst 445-5621 R. Kitch 696-3710 Columbf_a Fire _ Depar_tment 9-911-J. DeMers 445-2204 C. Kribbs 682-3980 UM Hospital ' .ind Clinics - T. Seeger 875-8656 Ambulai H - 882-6003 J. Baskett 474-2046 or 9-911 e
. .D.i r,e,c,t,o,r,',s, ,0ff,i,c,e, ,,, , , , , ,
Phone flo. Walk-in 882-8091~ S. Morris T4Y-T2" . J. C. McKibben 442-6728 Uff _ Health _ Physics - (Office)' 882-7221 p l 6r.. Phi 1 Lee (Home) i 445-5275 Jamison Shotts-(Home)_ 474-2194 _, , ,0p,e,ra t i on s Rea c t o r Chemi s t,ry, , , , , , , , ,,, David Spate (Home) - 657-9450 Phone No. Phone No. , W. A. Meyer T4Y-7676 M. Glascock 7fT4~B791T 882-6211 MI) Ne_ws_ Bureau _ i C. Edwards 443-7529 J. Schuh .874-3086 See . Public Information 882-6214. i: T. Schoone 474-6416 V. Spate 657-9450 Procedure for other _ or R. Hultsch 442-6653 phone numbers. (SEP-7)' 882-9142 C. Anderson 696-5506 B. Bezenek 445-5680 State Emerge 7gency (SEMA) _ncy Mana,gement_- 314-751-27
' ~ ' - ~
G. Gunn 875-1162 fl. Tritschler 474-9388 _NR_C_, Region III
, 708-790-5500 ]
L. Foyto 446-0491 (ANI) 203-677-7305 J. Fruitt 474-0774 R. Hudson 875-0451 V. Jones 445-2543 H. Kilfoil 474-6285 P. Neel 442-8693 H. Randolph 442-5315 P. Walker 445-8077 M. Wallis 443-8764 T. Warner 816-882-6740 , B. Warren 445-2204 . Rev. 12/08/89 App'd _ M\_ g ., -- ' =9w w--- % m ,y -m1 - ..w- i. ., 4 g ,sg.--.g, ..s e
. _ . . - - _., gJJ . _,
'1f, :t
., f' 1' w
', .J , WOR _KSHEETJ Cf.
CONTENT ~OF:IN!TIAL/ FOLLOWUP' NWi PE55A6Es To , ~
, 1 'T}{E NRC .
i , RIGT0VT11; 708_-_79_0-550_0_ ).. 1.: = Name, title, telephone number of caller,(882-4211. or 874-4119) Name'&
Title:
~ ~ ~
~ ' '
Telephone Number {" ; of Caller ____ _ __ _ l i Location of emergency er j M 2. Description of emergency event' and emergency clas',.. (NOTIFICATION 0F B- UNUSUAL EVENT, ALERT, or SITE AREA EMERGrlCY) r f g l Lg:-
- 3. Date~and time of event initiation. {
Date: _ _ _ _ _ _ _ _ _ _ Time: __ _ ,,,_ ___
- 4. Type of expected or actual release' with estimated exposure time. -
LI 1 Quantity of radionuclides released or expected to be teleased. g 5.
- g. ,
k
- 6. Impact of releases and recommended emergency actions.
- I:
' EMERGENCY DIRECTOR AUTHORIZATION TO CALL
-Rev.12/08/89 App'd M r . "
- ~ - --' - ~ ~ -
y .u;;v :<; v;:~ r
?-
i .
. MURR STIE EMERGENCY PROCEDURES AND I, FACIIJ1Y EMERGENCY PROCEDURES .."' Revision Number 5 Revision Date: March 22,1990 ; ; Section Page Number - Numher .
EMERGENCY CALL LIST FEP-1 1=
-c F E P 'I 2 FEP-1 3 FEP-1 4 FEP-1 5*
- FEP- 1 6* Orig.
'f FEP-1 7* Orig.
FEP-3 1
- Dated 5/22/90 in error.
I LI f g; E II 6 s h1 I h
. - = - ._-
suse. a us
. mas
~
sum. amm. aus mum u aus : sus mum. sum man sus- ins - EMERGENCY PROCEDURE EMERGENCY CALL LIST Health Physics Operations Emergency Support Organizations
" ~~
Ph~onle No. Phone No. Phone No. S. Langhorst 442-3534 M. Evans 698-2450. UNC Police 882-7201 R. Stevens 442-2539 K. Beamer 682-5499 J. Ernst 445-5621 R. Kitch 696-3710 Colunbia Fire Department 9-911' J. DeMers 443-4938 ] C. Kribbs 682-3980 UM Hospital and Clinics T. Seeger 875-8656 Ambulance 882-6003 J. Baskett 474-2046 or 9-911L Director's Office Phone No. Walk-in 882-8091 S. Morris 445-4217 . J. C. McKibben 442-6728 UMC Health Physics (Office) 882-7221 Dr. Phil Lee (Home) 445-5275 Jamieson Shotts (Home) :474-2194
,,,_0p,erations Reactor Chenistr_v David Spate (Home) '657-9450' Ph_one No. _ Phone No~._
W. A. Meyer 442-7675 .M. Glascock ___424-8390 _)tI. News. Bureau _ - y 882.6211 _ _ m See Publ ic Information ~~'-~ - 882-6214' C. Edwards 443-7529 J. Schuh '874-3086' T. Schoone 443-8862 ] V. Spate 657-9450 Procedure for other or' R. Hultsch 442-6653 phone numbers.'(SEP-7) ~ 882-9142 C. Anderson 696-5506 State Emergency Management B. Bezenek 445-5680
' ""' lagentr-(5tnA)-
314-751-2748 - G. Gur.n 875-1162 H. Tritschler. 474-9388 NRC, Region III 708-790-5500' L. Foyto 416-0491 (ANI) 203-677-7305 J. Fruits 4/4-0774 R. ..ud on 875-0451 V. Jo.. 1 445-2543 Kil...' 449-2524 ] .) T Neel 442-8693 H. Mc:.doigh 442-5315 R. Watker 445-8077 M. Wallis 443-8764 T. Warner 816-882-6740
- 8. Warren 443-4938 ]
Rev. 3/22/90 App'd tempfh
,y m -- - -
u i:,it . , R f. Paget1 ofa7.- ]: ;
, 'FEP-1l FACILITY E'/'.2% TION PROCEDURE - l 4
E NOTE: 'An assessment of ~offsite radiological consequences shall be determined. - E , J This -assessment' may ' require escalating emergency. response to 'a site emergency-procedure (UNUSUAL EVENT, ALERT or SITE AREA EMERGENCY).- ENTRY CONDITIONS: ,, j ! 1. The Facility Evacuation alarm is actuated manually-from two locations: (a) the reactor control room, =and (b) the lobby control center. h 2. Situations that may warrant FACILITY EVACUATION include: o . (a) ' Security emergr~.cies,. such as a bomb threat. l . (b)- A major facility fire. i 1
- (c) ~ 'Whenever airborne radioactivity is expected to- exceed 5 MPC throughout the-facility. j
.l (d) This procedure may be used as part of a Site- Emergency Procedure-(SEP).
(e) _0ther conditions occur that the Shift Supervisor determines warrant _ personnel evacuation from the facility. ' I AUTOMATIC ACTIONS: The following events result from a Facility eva ;uation alarm: )
~1. The reactor scrams. 1
- 2. The containment ventilation system isolation doors close.
- 3. The containment exhaust isolation valves close.
- 4. . The facility hor"s sound. l 4 q L 5. The flashing red light exterior to the containnent personnel l airlock door is energized. l 1
\ ; l l Rev. 3/22/90. App'd M l i m 1 1
- 1 l
<FEP-1 (Cont'd) Page?2lof.'7 ]!
I. PERSONNEL.= WITH PREASSIGNED TASKS (Facility Emergency Organization Members). A. IMMEDIATE ACTIONS: The responsibility for the overall direction in the event of an emergency shall rest with the EMERGENCY DIRECTOR.
~
! 'In the event of a Facility evacuation during normal working hours,- the - following people shall' report to the reactor lobby: the Facility . Director, Associate Director; Reactor Manager, Manager of Reactor Health
. . Physics, Machine and Electronics Shop Supervisors, Duty Shift ~ supervisor, and a representative of Reactor Chemistry.
i The responsibility for EMERGENCY DIRECTOR shall be assumed. The EMERGENCY DIRECTOR shall ascertain the availability of personnel req, tired to execute the emergency' plan and shall appoint an EMERGENCY C0ORDINATOR. He shall' investigate the cause of the alarm and the magnitude of the incident, and shall direct those activities necessary to correct.the emergency situation. After the emergency is terminated, he shall direct r. the procedures necessary to restore normal operation. I The EMERGENCY C0ORDINATOR shall ascertain that the reactor containment building, the Facility laboratories, and the mechanical equipment room, and below grade areas, have been vacated and secured. He will have the g< laboratory ventilation fans secured. He shall maintain a roster of all g persons released from the site by-the EMERGENCY DIRECTOR. If.the pneumatic-blower system was in use during the emergency, he shall insure that the samples being irradiated are returned to the laboratory and then have the I blowers secured at the local lighting panel- (#32). He shall insure a record of the events following the emergency is maintained. The DUTY OPERATOR shali perform or have performed the following tasks before leaving containment: (Da NOT attempt to correct any abnormalities at this time.) A. Verify that the reactor has scrammed as indicated by the i nstrumentation. [ B. Verify that all shim rods hace bottomed as indicated by the console lights.
;l C. Verify that the containment has sealed as indicated by the ventilation door and the exhaust valve lights.
Rev. 3/22/90 App'd W i l 1
,1
'FEP-l'(Cont'd) - s Page.3~of 7 ]'
D. Ensure.all personnel:are cleared from all _ levels of the contain-ment building -and exit via personnel airlock ~ doors. I 1 He shall report to the EMERGENCY;C0ORDINATOR and advise him of the status-of the reactor. I MANAGER OF HEALTH PHYSICS shall proceed to the lobby control center and establish the radiation-safe condition of the area. He shall estab-lish a' hot-cold change area, assemble and prepare for use special Health B' - Physics equipment, and perform radiation and contamination surveys. He 5: shall evaluate the extent of radioactive contanination and/or radiation exposure received by personnel in the Facility at the time of the inci- ,- -dent. 'He~ shall advise the EMERGENCY DIRECTOR of measures to be taken to control and to clean up radioactive contamination which may have resulted from the incident. I The EMERGENCY DIRECTOR shall appoint a COMMUNICATOR to notify auxi' iary organizations which have been made aware of these emergency procedures and perform other communicative functions required. The following telephone numbers may be of assistance in the -performance of these duties:- University Police / Watchman's Office, UMC 882-7201 Radiation Safety Office, UMC 882-7221 Dr. Philip Lee, 2 Research Park Dev. Bldg. Emergency Room, VM Hospital & Clinics, UMC 882-8091 NOTE: When determined appropriate by the EMERGENCY DIRECTOR,
-- the evacuation horns may be silenced by opening breaker 15.on the emergency lighting panel located in the north inner corridor next to the emergency power transfer switch.
I l I I Rev. 3/22/90 App'd u$$ g v
)
[ 1 FEP-II(Cont'd). Page:4 of'7: 1 II. PERSONNELWITHOUTPREASSIGNEDTASKS$ _ (Staff other than Facility, Emergency Organization members) A.- IMMEDIATE ACTIONS:- 1.- Upon' hearing the evacuation alarm, personnel shall- proceed to points I- beyond .the area bounded by the outer perimeter of the reactor labora . tory building.
- I -2. TOUR GUIDES shall be responsible for the safe evacuation of visitors in their charge from the Facility in accordance with the evacuation -
I routes in this 71SITORS shall be monitored by' Health. Physics Technicians asl plan.per HP '20~F6T6re being released to leave the site.'
-3 3. . EXPERIMENTERS who. are conducting experiments in the containment-Eg area shall render their experimental apparatus safe for unattended operation. They-shall be responsible for the safe evacuation of visitors in-their charge from the facility in accordance with~ the evacuation routes in this plan. -
j 4. EVACUATION ROUTES (See the map of the routes on page 6.): B A. All personnel within the containment building will exit the containment building and proceed through the east door of the laboratory building and then go to the: upwind parking lot, a -B. ~ All laboratory personnel, support personnel,. and guests exterior
.to the containment building will leave the facility through the
;' nearest exit (north, east, or south doors) and then proceed to iI the upwind parking lot.
Once outside,. personnel shall note the wind direction indicator 5.
) at the top of the containment building east tower and proceed to the upwind parking lot.
~;\
B. SUBSEQUENT ACTIONS:
;g. 1. All staff aersonnel shall remain on standby, unless released by the EMERGEiCY DIRECTOR, to provide the special services that- may
_ g! ' be required to testore normal operation. ,
- 2. All staff personnel shall be monitored by Health Physics Technicians as per procedure HP-20 before being released to leave the site.
* .A roster of all released personnel will be maintained by the 3. ]g EMERGENCY CO M INATOR.
_- u .- 7 an g 'Rev.-3/22/90 App'd (#41h 1
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EVACUATION ROUTES, RESEARCH REACTOR FACILITY FIGURE FEP-1 Rev,E /22/90 App'd tgaM l s
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Page 1 of~1 FEP-3 FIRE PROCEDURE _
- 1. - Any,ir.dividual ' discovering fire shall notify reactor control' (#13) of fire, giving nature and location of fire. The Shift Supervisor will activate the FACILITY EMERGENCY ORGANIZATION by page system and provide warning to stay-clear.of. fire location.
- 2. SHIFT SUPERVISOR will call (9-911) to notify Columbia Fire Department.
B .
-3. EMERGENCY DIRECTOR.will investigate the fire and determine steps to minimize hazard to both personnel and property.
NOTE: An. assessment of offsite radiological consequences shall be determined. This assessment may require escalating emergency I' response to a site emergency proceoure (UnusJai Event, Alert).
- 4. The EMERGENCY DIRECTOR may contact the MU News Bureau to handle public
, Il information, if appropriate.
. 5. If the fire cannot be put out immediately with local fire extinguishers - the reactor WILL be shutdown to focus on fire.
.-' 6. Secure EF-13 and EF-14.
- 7. Secure ventilation supply ara exhaust fans and close all fire doors.
- 8. _ If the fire is in containment and cannot be immediately brought under control, _]
initiate reactor isolation. If fire is in laboratory building and cannot be ] immediately brought under control, initiate Facility evacuation. ]-
- 9. The EMERGENCY COORDINATOR or EMERGENCY DIRECTOR should contact the Fire ]
.g1 ~ Department outside of the Facility and stay in contact with the INCIDENT ]
;g; COMMANDER to coordinate fire fighting and life saving efforts. ]
]
=
. NOTE: The Fire Department INCIDENT COMMANDER coordinates all Fire Department /
medical assistance personnel. The EMERGENCY DIRECTOR or his ]- representative should meet the first fire truck on the scene outside of ]
- th- 'acility. The first ambulance at the emergency scene is designated ]
'3 fer triage (sorting and allocation of treatment by injury priority) and ]
j will not provide transport fcr injured personnel. ] Rev. 3/22/90 App'o h !
.i
- .. . . . . . . . . . . . g
._ ........u_ -. . .
I SECTION III t REVISIONS 'IO THE HAZARDS
SUMMARY
REPORT.-
'f 1 July 1989 through 30 June 1990 ~
HAZARDS
SUMMARY
REPORT (original July 1,1965)
- 1. Original Hazards Summary, Section 3 Cooling Tower Below Grade Plan, Figure 3.9 Delete original Figure 3.9 and replace with new Figure 3.9 (See page
=
-III-2), revised due to removal of unused water softeners.
I 2. Original Hazards Summary, Section 7.1.4 Emergency Power System (pp 7-2,7-3,7-4} 2 Delete Section 7.1.4 of original Hazards Summary and replace with the following: 7.1.4 EMERGENCY- POWER SYSTEM Attached to the southwest corner of the reactor laboratory building is an addition that houses the emergency diesel generator and
;l provides space for future addition of a 1250 KW substation. - The emergency generator is a 275 KN/ diesel engine driven unit.
-g; Operation of the engine and generator is automatic. It starts one B second following failure of normal power. After reaching rated
- (
voltage and frequency, the unit will automatically assume the-
.! emergenc7 electrical load. Upon restoration of the normal electricrJ power source, the emergency electrical load will be "y
- automatically shifted after an adjustable delay time and the engine-
;g will be stopped after an additional adjustable time delay. *gE ..
The emergency generator (EG) is powered by an 855 cu. in., 395 h.p., diesel unit with a direct injection fuel system. The diesel EG is sized to meet current and anticipated loads with an excess j{ capacity approaching 50% for future load additions. The unit is
, designed to assume the emergency load within seven seconds of a cold start.
l The generator is rated for an output of 344 KVA (275 KW at 0.8
, PF), 277/480 volt, three phase, 60 cycles. The emergency power
_ generator will provide for the electrical requirements of the j following systems: i 111- 1
~
- l
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2 amuw tues i Figure 3.9 Cooling Tower Below Grade Plan
.(
III-2
-i .
~
I [4 r y.* , (1) Reactor Control Room Instrumentation-(2) Personnel Entry Doors and Controls ;
<c (3) Supply and Exhaust Air Doors and Centrols
~
- (4)- Facility Exhaust Fans (EF-13 and EF-14) k (5) Emergency Air Compressor
. (6) Evacuation / Isolation Alarm System (7) Fan Failure Warning Light System (8) Communication and Paging System
- (9) ' Exit Signs (10) Isolated Lights I (11) Stairway Lighting (12) Diesel Generator Electrical Controls +
(13) Offgas Stack Monitor
'f " Die system will be tested once per week for 30 minutes to assure operability.
1 Periodic maintenance will be performed according to manufacturer's recommendations.
- 3. Original Hazards Summary, Section 7.2.3 General Watersofteners.
- c. , Delete this section in its entirety. This system has been removed. I p i<
LI o '~ III-3 l
L I m.1 i l .4. Original Hazards Summary Report,.Sectil17, Paragraph 7.2.7.. Ventilation and- Air Treatments (p, 7-lTV Delete original paragraph and revisions submitted in 1974 Annual Report. Replace with the following paragraphs: The Research Reactor Facility building complex is totally air conditioned. The building air intakes are located on the north and !h. south faces of the east reactor containment building tower and include two roof top air handlers (RTAH) in the laboratory building g roof lone mid way on each 'of use north and south corridors).. -W: Building air is exhausted through a stack in the west tower. Air from the laboratory fume hoods is passed through-a system of absolute filters prior to being mixed with reactor containment building exhaust air and passes out of the building through the stack.in the west tower. Reactor containment building air that is discharged to the atmosphere is thermal column cooling air, beamport ventilation air, air which is drawn from the surface of the pool and exhaust from
.l the film irradiator shield box Reactor containment building exhaust air is mixed with and diluted by the laboratory building exhaust air. The reactor cooling equipment room ventilation air 1
and the pneumatic tube system exhaust air pass through filters and ' then also exhaust through the building stack.
- 5. Hazard Summary Report, Addendum 1. Section 3.8 Delete Section 3.8. including Figitre 3.8.1 and replace with the.following:
~
The emergency power system is driven by a water cooled Cummins, six cylinder, turbocharged diesel engine, it is provided with a 270
<g gallon skid mounted diesel fuel storage tank and a mechanically g; driven fuel injection system. It is capable of assuming full load from a cold start in seven seconds. A 24 volt, nickel-cadmium storage battery is used for the EG starting system.
111 4
-e ,
.. , w: - -
- =.. :. .
;h 8 I
m Attached to the diesel engine is a four pole ~ generator equipped' Jg with a brushless permanent magnet exciter. It produces 60 cycle.
- l. 3 277/480 volt, 3 phase power and has a continuous standby capacity; g of 275 KW. - he design of the exciter and regulator provides for J
' voltage regulation of better than plus or minus 2%. Stable -
generator output voltage and frequency are established within two
. seconds after the transition between no load and full load- ,
-lf. conditions. -
J :El ne automatic transfer switch (ATS) is equipped with an adjustable
.5 to 3 second delay on starting, preventing plant operation on ;
instantaneous line failures, and an adjustable O to 25 minute delay ; l [ ' on retransfer to ' commercial power._ Incorporated in the unit is a . H static type dual rate float / equalizer charger _with automatic and manual charge contrcl to maintain the startup battery fully charged.- The emergency bus is routed through the automatic transfer switch i l- to an Emergency Distribution Panel (CTR-1) shown on Figure 3.8.1.- l This. distribution Panel feeds the following emergency electrical loads: I t LI
-1 i
- i 1
g: . m g. III-5 1 I j __r _ _ 1 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ - . - _ __ __. . _ _
- m. g m eg m -g m g. g g m :g ' g-~ > m:' m 3- m 23 ;
NORTHTOWER ~ "*"
,=x lililil q uPs t_mac BYPASS 15KVA "*
NOMER ,' DIESEL SUBSTATION INNERCORRIDOR " " WESTTOWER
- GENERATOR NORTH SIDE P2 "
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,pg 120VAC D ~- 6. ~~w DIST.
PANEL WELDING SHOP ELECTRICAL DISTRIBUTION nevisions stocxonnoRAu wo__ i... c - o . RESEARCHREACTORFACUTY . ___ UNIVERSHYOF RESSOURI l . onewsav = APPRMED j-
, Shtt:I ND. 1 G1 UWG. NO. 2272 '[
Figure 3.8.1 j
. ~~. - _- _________________ __ _____ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ ___-_-
; i
- 1) Exhaust Fan EF-13 J
- E . 2) Exhaust Fan EF-14
- B --
- 3) Diesel Room Distribution Panel which provides control power l' for the EG room ventilation system.
_ g1 !
- 4) Emergency Power Panel which feeds through a transformer and
. : 5. distribution panel to supply exit lights, stairway. lights, fan failure .. alarm, intercommunications system, and the evacuation alarms.
.( This emergency power panel also feeds the emergency air complessor, the motorized isolation doors, the truck entry door
; and the pedestrian entry doors.
1 5) 120 volt distribution panel via either and uninterruptible power i supply or a line conditioner. This distribution panel provides
-1 the following loads:
a) Stack Offgas Monitor g; b)- Reactor Control Power (control rods, rod run-in, safety JB system, and-Servo ampitfier)
~f 'c) Annucciator Panel d) Area radiation monitoring system e) Neutron and Process Monitoring Instruments The generator will run for approximately 30 minutes weekly under no load conditions. The generator is load tested on at least-a semi-4
! annual interval.
-; - 6. Ha::ards Summary Report, Addendum 1, Section 3.22, (pp 101-102)
_, Delete Section 3.22 in its entirety and replace with the following:
" Submit drawings showing the general arrangement of the
.g; ventilation systems and associated dampers and controls for the
_ E containment and laboratory areas." ]
- a. .
+
III-7
+g
_ . . . . . ~ 7 _
, 7 _ _ , ,
if l , c . . 1 Figures 3.22,1 and 3.22.2 illustrate the ventilation system for the' laboratory and the containment building. 4 i
~
l All fresh air for both the' laboratory and the containment building. .j enters through dampers on the north and south faces of the east c g- tower and through the dampers of each RTAH unit. Fresh air j u- entering through the north and south dampers passes into receiving
- plenums and through steam preheat coils.
h The fresh air then passes through a dust filter, moving on throu'gh Q. supply fan No.1, (SF-1) heating and cooling coils, and finally into .' .: E- the double duct air distribudon system.
'. 3 Fresh air from the north and south RTAH's passes through chill water coils for air conditioning and second".ty system reactor waste '
heat :: oils for heating, and is distributed via ceiling grills.in the north and south c'orridors. Fresh air for the containment building passes up from the receiving plenum and is mixed with containment: building return air. Containment building return air, driven by return fan RF-2, enters the east tower through the motorized isolation door No. 505. The mixed return plus fresh air passes
- (
through a dust filter, cooling coils, heating coils, and motorized isolation door No. 504 to supply fan SF-2. From supply fan (SF-2) I
]
the air is distributed throughout the containment building. laboratory building air and contalmnent building air are never l mixed. The supply and return An pairs are interlocked so that if l SF-2 is off, so is RF-2, however, RF-2 may be off with SF-2 on. If either of the motorized Isolation :loors (504 or 505) are closed. SF-2 and RF-2 are both off. lg- Exhaust air from both the containment and the laboratory building L. enters the atmosphere through either exhaust fan EF-13 or 14 and I then through the exhaust stack located in the west tower. Either . 7h one or the other (EF-13 or EF-14) exhaust fan is on at all times. The other fan is a standby. Failure of the on-line fan automatically
-g activates the standby and also activates a warning light in: (a) the ca reactor control room; and (b) in the facility lobby. Failure of both fans activates an alarm buzzer only in Reactor control room.
I. III-8 l
l
;c
. . }:
g
' Laboratory exhaust air is picked up _at the fume hoods, passed -
, through absolute filters, and delivered to EF-13 or 14. Exhaust air; < l l - from the mechanical-equipment room is delivered to EF-13 or 14' ! L .through activated charcoal and absolute filters. Containment building exhaust air enters EF-13 or 14 through two (2) quick-closing isolation valves. This exhaust air is picked up from :; beamport experiments storage ports, beamports, the thermal column, the Nuclepore film shield box, and the pool surface air _ if sweep. , 1 - 7. Hazards Summary Report, Addendum 3, Section 2.3, Page 20, Figure :. p l: 2.2 ' H. E Delete original Figure 2.2 and replace with new Figure 2.2. This drawing bF was revised to reflect improvements in the air conditioning system and . removal of a Chromate mixing _ tank. 1 l 8. Haiard Summary Report, Addendum 3, section 2.4 Figure 2.3 ' Delete originel Figure 2.3 and replace with updated Figure 2.3. l v I g ,I;.. I III-9
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- 9. Hazards Summary Report,' Addendum 3, Section 5.3r J' Efiluents '(pp. _ 193-194):
Add new Section 5.3.3'- I, EVALUATION OF ENVIRONMENTAL IMPACT OF INCREASED STACK RELEASE FLOW RATE,' _ (see attachment, pp. Al-A18)'-- I I I. I. It I g
;gL 4 -
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g 111- 1 2
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l- a EVALUATION OF ENVIRONMENTAL IMPACT. OF INCREASED STACK RELEASE FLOW RATE' i Stack release limits set for MURR in Technical Specifica-tion (1) Number 3.7: " Facility Gaseous and Particulate Radio-active Release" are-based on activity concentrations. An increase in stack flow rate affects the total allowable release of activity, and thus this evaluation is made to
- I l~
. assess the environmental impact the increase will have on the nearest resident and on the population surrounding the MURR.
The change in stack height and exhaust exit path is also con-1 . sidered. The safety significance of the impact is discussed in relation'to background radiation and in relation to a previous environmental impact appraisal made by NRC. (2) I Data and lamustptiana I The data and calculations in Table 3 describe the physi-cal =information~of the stack release point. principal-isotope released in gaseous effluents from MURR. Argon-41 is the The Technical Specification limit for Ar-41 release is 350 I t times.the MPC listed in Appendix B, Table II,' Column I of
-10CFR20,'or:
l Q = 350 x MPC x flowrate
=
(350)-(4x10-8 p ci/ml) (36500 f t3 / min) (2.831 x 104ml/f t3)- i
= (1.4 x 104 pCi/ min) (1x10-6 C1/ C1) (1 min /60sec)
= 2. 4 x 10-4 Ci/sec In the previous environmental assessment, (2) the NRC used meteorological data collected at the Callaway Plant, located
'I. near Fulton. These data were collected between May 5, 1973 and May 4, 1975, and were judged by the NRC to be " reasonably representative of long-term conditions expected at the MURR ,
site." This current assessment utilizes meteorological data gathered in Columbia, MO from 1960 to 1969. (3) The Columbia data was judged to be more appropriate for use in assessing airborne releases from MURR because of the longer data period and the proximity of the data site to MURR. Table 2 lists E wind data (stability, class, speed and frequency) for each of ' the sixteen campus points. A-1
~
i
' I } 'l I ,- Table 1 '!
Physical Information for Stack-Release Point
^
Elevation above' sea level = 687 feet Diameter = 40 inches-New Max flowrate = 36500 ft /3 min P
"I- Area cross section = ar2 l
. g[ 40-inehan '2 i 212 inches /ft i i
'~ I' = 8.73 2t2 ; Air velocity (v) = 3E500 ft3/ min 0.304 m/ft . I "4" 60 sec 2 i 8.73 ft
= 21.2 m/Sec jg :.
IH < ll 1
, , m=: . w. ~ . : ::. '
; 3..
t I 1 L I Meteorological Data--Columbia, MO (1960-1969) (3) 1 [ I Stability class information NNE Nind Class (*) 4 Class (b) speed (C)- % NNE(d) 4's(*) }
~
im/mac) wind r=nsnh ;
- A. 0.4 2 . "* 3.4 1.4e-04 ,
B 4.7 2.a 2.7 1.3e "El C 11.5 4.0 3.5 .4.0e-03
?g D 53.6 '5.7 .4.2 2.3e-02 E 17.6 3.8 3.2 5.6e-03; r 12.2 2.4 4.9 6.0e-03 -W.
B: 5 Stability class information NE L Wind 1
/g g Clas s (a) % Class (b) speed (c) % NE(d) % 's (*) q (m/ nee) wind enmh.
.. A 0.4 2.1 1.7 6.8e-05: -
i B 4.7 2.7 2.6 1.2e-03 C 11.5 3.7 2.7 3.le-03
,E~, D 53.6 5.2 3.9 2.1e-02 3 E .17,6 -3. 6 2.8 4.9e-03 F 12.2 2.5 4.9 6.0e-03 l
Stability class information ENE ! Wind '!g Class (a). % Class (b) speed (c) % ENE(d) % 's (e)- g (m/see) wind enmb. L A 0.4 2.0 7.8 3.le-04 L ; B 4.7 2.8 5.1 2.4e-03 t : C 11.5 3.9 4.3 4.9e-03
- - . D 53.6 4.9 4.7 2.5e-02 L
l E 17.6 3.4 4.2 7.4e-03 F 12.2 2.5 6.8 -8.3e-03 "I A3
x
, ' Stability class information E Wind- 1 Class (*)' 4 Class (b)
~
speed (c) % E(d): 4 's (*) - l s . Im/neci wind en=h. , q A 0.4 '2 . 0 4.3 1.7e-04 3 4.7 2.9 5.3
~
2.5e-03 . C 11.5 3.8 4.4 5.le-03 I D 53.6 4.9 4.4 2.4e , E 17.6 3.5 5.0 8.8e-03 F 12.2 2.5 7.9 9.6e-03 Stability class information- ESE II ;5 Clas s ja) - Wind
% Class (b)' speed (c) % ESE(d) % 's (*) i (m/sec) wind comb.
A 0.4 2.0 3.4 1.4e-04 I' B C 11.5 4.7 2.9 3.9 4.7 4.8 2.2e-03 5.5e-03 D 53.6 5.3 6.1 3.3e-02 l ' Wl- E F 17.6 4.0 6,1 1.le-02 12.2 2.6 4.5 5.5e-03 IL z-' Stability class information SE Wind l -. Clas s (a) % Class (b) speed (c) % SE(d) % 's (*) li (m / =aM wind enmb. l A 0.4 2.2 .2.6 1.0e-04.
- B 4.7 2.9 4.6 2.2c-03 ll C 11.5 4.1 6.4 7.4e-03 D 53.6 5.7 7.6 4.2e-02 E 17.6 4.1 8.2 1.4e-02 F ~ 12.2 2.5 4.3 5.2e-03 g-I
.I-A4
I 3 Stability' class information .l g 333-Wind t Clas s (*) % ' Class (b) speed (C) % SSE(d)- % 's (*) ) (m/mac) wind rn=h. l A 0.4 2.3 4.3 1.7e-04 ! B 4.7 3.0 6.5 3.le-03' LI C D. 11.5 53.6 4 .1 ' 5.6 8.7 9.3 1.0e-02 5.0e-02 --1 E 17.6 4.1 12.0 2.le-02 F 12.2 2.7 7.2- 8.8e-03 LI Stability class information S Wind .. l Class f a) % Class (b) speed (c) g s(d) 4 's (*) . l (m/sec) wind comb. I A 0.4 2.1 6.0 .2.4e-04 B 4.7 3.0 10.8- 5 le-03 l
- g. C 11.5 4.2 14.4 1.7e-02 I g- D 53.6 5.6 11.8- 6.3e-02 )
E 17.6 4.0 17.6 3.le-02 12.2 F 2.6 12.0 1.5e-02 Stability class information SSW_ Wind
- h. . Class (a) % Class (b) Speed (c) g ssw(d) g e 3 (e)
'- '4 (m/sec) wind- comb. ,.
'l- m A
B 0.4 4.7 2.4 3.1 6.0 8.6 2.4e-04 4.0e-03 C 11.5 4.1 9.7 1.le-02 D 53.6 5.6 5.5 2.9e-02 i.IL E 17.6- 3.9 7.4 1.3e-02 F 12.2 2.6 6.3 7.7e-03 I A-5
7 , ; I 1 L Stability class information SW Wind Class (*) %. Class (b) speed (c) g gw(d) % 's (*) !' f m/ mme) wind eamh. 0.4 5.2 A 1.8 2.le-04 ^ B 4.7 3.0 9 '. 2 '4'.3e-03 i C 11.5 4.1 7.5 8.6e-03 1 : D 53.6 5.4 3.5 1.9e-02 E 17'6
. 3.9 4.3- 7.6e 03 ,
F 12.2- 2.5 6.0 7.3e-03 3
.IL Stability class information WSW
!EL Wind
~Et- Class (*) - % Class (b) speed (c) % WSW(d) % 's (*)
im/sec) wind enmb. L A 0.4 2.2 6.0 2.4e-04 l B 4.7 3.0 10.8 5.le-03 ,
.,. C 11.5 4.3- 9.0 1.0e-02 D 53.6 5.9 4.9 2.6e-02 t J .
E 17.6 3.9 5.7 1.0e-02 F 12.2 2.5 5.9 7.2e-03 Stability class information W t- s Wind gis (*) Class (a) % Class (b) speed (c) g w(d) (m/sec) wind- comb.
-E ^ 04 1.8 3.4 1.4e-04 l ,3 .B 4.7 2.8 6.7 3 le-03 L C 11.5' 3.9 6.2 7.le-03 l- D 53.6 6.0 4.7 2.5e-02 E 17.6 3.7 5.3 9.3e-03 12.2 r 2.5 6.1 7.4e-03 b
I A6 I
g --mgryc.-
;.-- ; -- ;- 7-y i
~4 ..
l Stability class.information WNW-i< Class (*)' % Class (b) Wind speed (c) g www(d) ge s te)- (m/mmel wind enmh. A' O.4 -2.1 .4.3 1.7e B 4.7 2.8 -5.4 2.5e-03 I; C D 11.5-53.6 4.3 6.7 5.1 7.9 5.9e-03' 4.2e
- E 17.6 4.0 5.5 9.7e-03 F 12.2 2.5 5.0 6.le-03 I Stability class information NN-lE Wind L5 Class (*) 4 Class (b) speed (c) g ww(d) - ge s (*) ,
(m/nec) wind comb. '
. A 0.4 2.2 4.3 1.7e-04 B 4.7 2.0 4.4 2.le-03 ,
n . C 11.5 ..a 4.7 5.4e-03 l- : D 53.6- 7.1 8.8 4.7e-02 E 17.6 4.2 5.1 9.0e-03 F 12.2 2.5 3.6 4.4e-03 1 Strbility class information NNW-Wind Class (*) % Class (b) speed (C) % NNWid) % 's (*) nI. l im/seci wind comb.
- 3 A 0.4 2 .'3 1.7 6;8e-05
E- B 4.7 2.7 2.9 1.4e-03 C 11.5 4.1 3.0 3.5e-03 lJ D 53.6 6.6 5.8 3.le-02 E 17.6 4.0 3.6 6.3e-03 F 12.2 2.4 3.0 3.7e-03 C
.- A7 m_ _ - - - _ - .. . . - _ __. -
_, 7 I
~
1
.Ib . :
Stab'ility c' lass information' N-Wind- I Class (*) ' n Class (b) speed (c) g y(d) ge s (*) im/m s wi n,i ra-w A' O.4 2.4 7.8 3'.le-04 1 B 4 . 'i 2.7 4.8 2.3e-03 iI C D E 11.5 53.6 17'.6 4.0 6.0
'3.8 4.8 6.2 4.0 5.5e-03 3.3e-02 7.0e-02 J
F 12.2 2.5 5.8 7.le-03 ,
=
(a) Stability class as defined by Pasquill's
- Categories .' (il' ,
-(b) Annual frequency distribution of stability ->
h . class for all directions, or the total probability of occurrence for that class. (c) Average wind speed for stability class and' wind i direction. (d) Annual frequency. distribution of wind direction for i the specific stability class, or the probability of. thel wind direction given that the stability class ,
', exists..
., (*) % 's comb . = (%-class /100) x (% NNE/100), or the joint i probability of the specific stability class and the l
specific direction occurring at the same t1me. l Example: A conditional probability is one'in which Lt
. the probability of the events depends:upon whether the other event has occurred (5) ,
J P ( A) = probability of Class A conditions = 0.4%. 1 - P (N/A) = probability of wind direction from N given-Class A conditions = 7.8%.
.P (AN) = probability of having Class ~A conditions and' wind direction from N.
I.. P ( AN) = P (A) P (N/A) = 3.1 x 10-4 s i o , It l' A8 I
- - ~ _ . _ _ _ .
I Listed in Table-~3Iaxe the_ equations used to calculate the Ar-41 concentration 1and dose, along with the associated gE assumptions used for each case, at:a distance, x,= downwind I from the stack release point. Calculations are based on-the Pasquill-Gifford Method of determining stack release concen-trations (effective stack height). Data for e n were obtained from= Ref. 4 and the DCF from Ref. 6. y and o m Tabla 3 Equations and Assumptions (1) Ef f active Stack Height W .(H): l 3 H=h+dM^ ip, i 1+ETi (Eq -1)- where,'h = actual height (m)
= difference in elevation from release point to downwind' site'of dose calculation i d = diameter.of release point (m) p = average wind speed for specific stability class (m/sec) y = exit velocity (m/sec)
AT = temperature difference between stack air and surrounding air
= assumed to be 0 T =' absolute' temperature of stack air-Therefore, I-- H=h+d Y- . (Eq. la) ,
ip, I (2) Concentration Calculation: E.= 1 exp lh + [ (Eq. 2) o ac y cip 2(ej 4 where, X = concentration at downwind site of dose calculation (pC1/ml or C1/m3)
- g~ Q = release rate (Ci/sec) ig- cy = lateral dispersion coefficient at downwind site of dose calculation (m)
_- oz = vertical dispersion coefficient at downwind site of dose calculation for specific stability class (m) p = average wind speed for specific stability I ., class (m/sec) y - distance from plume centerline (m) for maximum concentration, assume to be 0 H = effective stack height (m) l A9 I.
l } For-maximum-concentration: E- 1 exp ' 1Y , (Eq. 2a)
-Q 80 rO M-' , 2 lo s, .
Furtber, for case .of ground release , (H=0), cl E=' 1
. (Eq. 2b)'
Q 20 yOsk
- Considering decay, the equation becomes- .;
.%, , e-h (Eq. 2c) i Q- 2OyO M q s b where,'A = decay constant for Ar-41 (sec-1) t - time (sec)
" */E l:l (3) Annual-Dose Calculation (D):
i l D = DCF [ Xi (% comb)1 (Eq. 3) I 2 I i where,-DCF = dose conversion factor
= 8. 84 x 10-3 prem m3 for Ar-41
i = summation ovf a~fl. stability classes i l- (% comb)r = relative frequency for stability. class, i, and specific wind direction Ell g lg (I \ A10
l naaminum Individual nana To determine the maximum individual dose, the south wind 1 L , direction was chosen _as being:the most probable and annual , i doses determined at maximum release rate for two different -i distances:. 150 a north to the exclusion boundary,0) and 760 . m north to the nearest residence. Elevations for these two- [ l sites were estimated from a University..of Missouri topographi-cal' map-(shown in Fig. 1).- Data and the maximum calculated c dose; estimates for these sites are given in Table 4, with an l example' calculation _given in Table 5. The maximum average
. annual dose at 150 m was calculated as - 2' mrem /y and - 18 mrem /y and at 760 m. The difference in relative plume height-
;I ~
at these sites is what-leads to this difference in dose rates. Table 4 Maximum Average Annual Individual Dose FI: Location at_150 m Directly North Elevation at man height: 636 ft. l- Eff % Dose
- Class height o c .%/Q (pC1/ml) w/%'s
, (m) (mf (m) (s/m3) (Ci/m3) (mrem /y)
LI ' A 42 35' 23 3.6e-05 8.6e-09 O.0 B 31 25 15' 3.3e 7.9e-09 0.4
<I) C 25 19 11 2.5e-05 6.0e-09 0.9 L D 22 12 7 4.5e-06 1.le-09 0.6 E 26 9 5 1.9e-09 4.6e-13 0.0 F 35 6.6 3.2 2.7e-28 6.Se-32 0.0
' TOTAL 1.9 mrem /y
- E. Location at 760 m. directly North, g' Elevation at man height: 700 ft.
A 23 160 300 3.2e-06 7.8e-10 0.0 jI B 12 110 90 1.le-05 2.Se-09 0.1 C 6 81 50 1.9e-05 4.5e-09 0.7 D 3. 54 25 4.2e-05 1.0e-08 5.7 , 41 6.7
.I' .E F
7 15 30 18 11 1.0e-04 1.4e-04 2.5e-08 3.5e-08 4.5 TOTAL 17.7 mrem /y g A-l'l O _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ . _ _ _ _ _ . _ _ _ _ _ _ . _ _ . _ _ _ . _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - -= --
j l
'( , Table 5 Example Calculation L
l' g ' Distance: 760 m North - [g- Elevation: 700 ft. Class E: p = 4.0 m/sec ' cy = 41 m c = 18'm Effective Stack Height:~
^
H = 687 +61[1D.f 2 4.0i 21. 27 * * . 700 1"
; = 21 f t -
3.2808 ft
. = 6.5 m ,
1 l- Ar-41 Concentration: I. 1 exp dd 26 18i o - Q K (41) (18) (4 ) ,
, g, -;
- g. . = (1. 08 x .10-8) (0. 94 )
l -- = 1.0 x 10-d JAC-m 3 I . x=, 1.0 x 10-418C-m 3 d
'(2. 4 x 10**C17sec) i
= 2.4 x 10-an m3 an = 2.4 x 10-8pCi '
g ml Class.E occurs 17.6% of the time and of that time the wind blows fro:. the South 17.6% of time , j
%'s comb = l17. 6 M, = 0. 031 6 100 100<
12pci l Doset = 2.4 x 10-e n'l0.031) 8.84 x 10-3 mrem-m3' 10 pCi y 1 Cil i m1 3 i i
= 6.7 mrem /y I A12
_ m _. . 1 __ _ . _. . l Maximum Dannlation nana natinte
' Population' dose estimates were made assuming l ground release conditions. Population density data was generatedW- t
_using 1980 census data, 1985 update & n, and growth projec- , J : tions provided by City of Columbia oi .cials. Estimates for i population doses were based on the piajected 1990 population densities (See Table 6) . The max! um average annual dose was determined at the-center of each population zone, except for the 16 zones'at 0-1 miles. Because residences are no closer than 760 m, the midpoint was chosen at 0.75 miles (1200 m) from MURR. In LI' . L addition, radioactive decay was considered in these calcula- ; tions due-to the significant amount of time required for the
! plume to' move to these distances. Otherwise,_ calculations !I: were made as were the individual dose estimate calculations.
Data for c yand o, is given in Table 7, che summary of annual
.' ~ doses in Table 8, and the population dose estimate in Table
- 9. For the. population out to 10 miles, the maximum annual popula-tion dose is estimated to be 145 person-rem.
; Tahin 6 l
3' Projected 1990 Population Densities L3 (Number of People) Midpoint Distances'(m) Wind- . 1200 2400 4000 5600 7200 12000 Direction L NNE 238 437 368 315 262 206 l: NE 101 845 469 105 210 204-E- 1 ENE ' 132 534 449 440 76 305 l: E 94_ 1189 1270 3905 220- 315 L ESE 186 1138 2025 849 51 3850 SE 406 2096 1664 1021 474 402-SSE 354 2747 1676 542 428 920 L S 364 2649 2293 644 157 5750 1513 6135
~
SSW 1131 3163 1843 1404 2699 5137 2491 2387 1571 2877 E SW ; WSW 1997 4803 1067 1146 1055 5610 J
.- .W 49 1446 525 385 153 234 WNW- 52 592 1182 325 364 316 NW 36 126 644 222 103 315 288 665 229 30 154 210 ;I NNW-N 339 851 974 432 210 255 A-13
E :
' i- M 3
c y 's (top) and c,'s (botton) for - population distances and stability I - Stability Midpoint Distances (m) Class 1200 2400 4000 5600- 7200 12000 t A '220 400 620 900 1050 1800 800 5000 9700 14000 19000 33000
- B '170 310 480 690 820 1300 l 150 470 1100 2200 3300- 6600 '
C 130 220- 340 480 600- 900 1 75 130 'N.t e' 270. 320 500 1 :- ,
! D 80 140 220 300 400 610
# 34 53 72- 91 100 140 .
l-( 3 - E 60 110 170 220 300 460 , j3 23 40 50 60 70 84
< F 42 80 120
- 160 200- 300 14 22 30 33 40 48
;g ,
-I !
~
- I:
L, . A14 I
I 3.-
.i 7 !
s y M I 4 A Susanary of dose rate estimates- (arer./y) based on wind direction & distance Midpoint Distances.(m) ,
. Wind 1200 2400 4000 5600 7200 12000 ,
Direction L NNE 4.5 1.4 0.7 0.4 0.3 0.1 l1 .. NE 4.3 1.4 0.6 0.4 0.2' O.1 ENE 6.1 2.0 0.9 0.6 0.3 0.2. E 6.7 2.1 1.0 0.6 0.4 0.2 ESE 5.2 1.7 0.8 0.5 0.3 0 .1 ' : I SE 5.9 1.9 0.9~ 0.5 0.3 0 ~. 2
- SSE 8.4 2.7 1.3 0.8 0.5 0.2 13.0
.S 4.2 1.9 1.2- 0.7 0 . 3 --
SSW 6.3 2 '. 0 0.9 0.6 .0.4 :0.2
~
SW 5.1- 1.7 0.8 0.5 0.3 0.1 WSW 5.7 1.8 0.8 0.5 0.3 0.1
-W 5.6 1.8 0.8 0.5 0.3 0.1 0.1 WNW 5.5 1.8 0.8 0.5 0.3
! NW . 4.7 1.5 0.7 0.4 0.3 0.1 NNW , 3.7 1.2 0.6 0.3 0.2 0.1 !
.N 5.5 1.8 0.8 0.5 0.3 0.1 f
jg r . LI E A15
[ o i 3 ll .
. .i' 2&BInR 1 + ! EL Person-Rem Estimates (person-rem /y) t Midpoint Distances (m)
Wind 1200 2400. 4000- 5600 7200 12000 J Direction NNE 1.1 0.6 0.2 0.1 0.1 0.0 L NE 0.4 1.2 0.3 0.0 0.1. 0.0 l g, ENE 0.8 1.0 0.4 0.2 0.0 0.0 g E 0.6 2.5 1.3 2.4 0.1 0.1 ESE 1.0 1.9 1.6 0.4 0.0 0.5-SE- 2.4 4.0 1.5 0.6 0.2 0.1 I' S SSE SSW 3.0 4.7 7.2 11.1 7.5 6.5 2.1 4,5 1.7 0.4 0.8 0.8 0.2 0.1 0.5 0.2-1.9 1.0 .,
.SW 13.9 8.5 1.9 1.1 0.5 0.41 L WSW 11.3 8.7 0.9 0.6 0.3 0.8 l- W 0.3 2.6 0.4 0 . :2 0.0 0.0
!= WNW 0.3 1.1 1.0 0.2 0.1 0.0 L Nu O.2 0.2 0.5 0.1 0.0 0.0 l
- -~ NNW 1.1 0.8 0.1 0.0 0.0 0.0 N .1. 8 1.5 0.8 0.2 0.1 0.0-
- i. i Subtotals 50.0 59.8 19.2 8.2 2.4 5.3 l
L TOTAL -144.8 I I
- E -
I E A-16
u- i cammid-eatian af use==1 annentia=mi malammen i I -For.the past-five years, MURR has released - 1000 Ci/y. of Ar-41 with a stack ~ flowrate of - 16,500 ft 3/ min. Produc-Ltion of Ar-41 is expected to remain the amme, and so the average Ar-41 concentration is anticipated to be:
-I;
)
3.7E-6 (pC1/ml) 16500/36500 = 2E-6 pCi/ml which is - 13% of the Technic'al Specifications Limit. Because the dose estimates calculated thus far are propor- ' I tional to the total' amount of Ar-41 released, the dose esti- l mates for actual operating canditions are easily calculated i using the ration of the stack release flow rates (given Ar-41 ; production remains-constant). The actual operational dose ;
=
estimates-are: l Individual 9 150 m = 0.2 mrea/y l Individual 9 760 m = 2 mrem /yr Population to 10 miles = 15 person-rem connarisen of Risk f In theNo. Amendment Safety ) Evaluation 12,t2 an individual madelocated by the NRC at thein support nearest of resider:43 was estimated to receive an annual average total body dose of 13 mrem per year based on the 1977/78 release of 1925 Ci/y and 29 mrem /y for the. maximum estimate. In the same NRC evaluation, the population dose for implementing Amendment No. 12 was estimated to be 20 person-rem. Although ' assumptions, data, and conditions for calculation are not fully. described in the NRC Amendment No. 12, estimated doses are greater than those predicted by the current assessment, which utilizes more realistic model (effective stack height
; and stability class weighting) and better site-specific data (meteorological data and updated population densities). The NRC concluded "that_there would be no significant environmen-tal impact attributable" to an increase in stack release I"- limit to 350 MPC. With lower doses estimated for the current change in stack height and flowrate, it is also concluded that no significant environmental impac exists. The same conclusion applies to instantaneous release limits.
Another method of assessing risk from the estimated
;g: doses is to compare them to natural background dose rates. 'g. The average whole body dose to an individual in the US is 360 mrem /y . (9) The estimated doses in terms of % of natural backgrcund are:
I A17 I
Il !1 Maximum' Case? Normal-Operation-l' " Individcal 8 150 m- 0.5% :0,14. I' .IEdividual 9 750's Population 0.4% 54 0.6%
<0.1%
i I Variations of this-magnitude can be/found in annual dose for populations living in different areas of the US with no observable effects. + 1 conclusion i i The estimated dose rates calculated using improved methods and data were no greater'than those calculated from i 'a previous appraisals where impact was judged by the NRC to be g not significant in environmental impact. Therefore, there is - no significant. reduction in safety as the result of the changes'in the MURR stack release conditions. ll Rafarancea ! g (1) Appendix A: Technical Specifications for University of Missouri-Research Reactor Facility--Facility Operating License No. R-103. - I _> '. (2) NRC Amendr.ent No. 12 for R-103, July 5, 1979. lE (3) Callaway Environmental Report, Operational. License 1, Tables 2.3-19 and 2.3-20. L 3-1
. Stage, Vol.
(4) Cember, Herman, Introduction _ to Health PhysicM, Second , Edition, Pergamon Press, 1983, pp. 340-352. i D- '(5)- DeGroot, Morris H., Probability and Statistics, Addison- ,
-Wesley Publishing Company, Inc., 1975, pp. 49-50.
(6) Regulatory Guide 1.109: " Calculation of Annual Doses to Lg Man From Routine Releases of Reactor Effluents'for the Lg Purpose of. Evaluation Compliance with 10 CFR Part 50, L 1 Appendix I," Revision 1, October 1977, (7) NRC Amendment No. 8 for R-103, February 24, 1978. l (8) Environmental Report for Upgrade of MURR, 1987 (Draft). (9) NCRP Report No. 94: " Exposure of the Population in the United States and Canada from Natural Background Radiation," December 1987. LI 's A18
6
' 10. . Hazard Summary Report, Addendurr '5. Section 2.0 and Figure 2.1 Delete original Section 2.0 (pp. 2-14 and Figure 2.1) and replace with' the following:-
I SECTrlON 2.0- ANALYSIS OF A LOSS OF ELEURICAL POWER TO THE i MURR 2.1 Introduction ^ ',
'1his report contains an analysis of a complete loss of power at ~
the MURR, This implies a loss of commercial nower followed hv a failure of the emergency generator system. The L emergency generator system is described and the routine surveillance tests are outlined. Accident analyses will then be L presented for a complete loss of electrical power during a i period when the reactor is shut down. 2.2 Descrintion ,g Upon'a lo'ss of normal electrical power to the facility, the EG g- assumes the desired electrical loads. Drive power to the ' generator is provided by a Cuminins six cylinder, turbocharged l diesel engine. The engine is provided with a 270 gallon fuel storage tank and a mechanically driven fuel injection system. The EG is capable of assuming full load from a cold start in El 1 seven seconds. A 24 volt nickel-cadmium storage battery is ' used to start the EG. A static type dual rate float / equalizer charger automatically maintains the startup battery fully-LI charged. ( A four pole generator equipped with a brushless permanent magnet exciter produces 60 cycle 277/480 volt. 3-phase
.g service, and has a standby continuous load capacity of 275 KW, ,E! The design of the exciter and regulator provides voltage regulation of better than plus or minus 2%. Stable generator output voltage and frequency are established within two seconds after a transition from no loaa to full load conditions.
I 111- 1 3 I
ii
. _ =. _
Ey( I y y;. i ( An automatic transfer switch (ATS) selects the power source- ; 1
~ for the emergency electrical loads' fromLone of the' two inputs:- 1 1)- Commercial Power (i) City Power Plant, or
@h , (11) University Power Plant B, l 1 (2) Emergency Generator Power ne During normal operation, all loads are supplied from commer-
'l5-cial power. - Whenever a commercial power failure occurs for R i
greater than one second duration the engine starts, the- { nutomatic transfer swnch functions, and the EG assumes the load. Commercial power must be restored for a full ten ! minutes before the transfer switch functions to transfer the ! .- load to commercial power. ; We EG will continue to run five minutes after the load is transferred back-to commercial power in order to cool down ] the engine, i The EG and engine are located in a building addition on the -) L southwest corner of the laboratory building. He diesel generator room has local temperature controllers to maintain - room temperature above 55 F. The EG starting system is f ' designed to start the EG at temperatures as low as 32 F The 1 . operation of the temperature controllers is checked every four hours.by the operating staff during reactor operation. L c The emergency bus is routed through the automatic-transfer
~~
switch to an emergency distribution panel located on the wall , L in the north inner corridor of the laboratory building. This J: feed panel distributes power to the following circuits (see flg. , L u. 2.1.): I L (1) .Two circuits service reactor and laboratory exhaust fans 4 EF-13 and EF-14 located in the west tower. (2) One circuit services a 120 VAC distribution panel providing power for exit lights, stairway lights, fan failure E aiarm, intercommunication system, and the reactor B~ evacuation and isolation alarms. III-14 i
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o , l i (3) One circuit' services a 120 VAC distribution panel via either
~
an uninterruptible. power supply or line concitUoner. This j 7 ' distributiont panel provides power to the area radiation - l monitoring system, the annunciator control system, control .I room clock, all nuclear and process instrumentauon in the j control room. including control relay, solenoids, indications j
-of primary, pool and other valve posiuons, control rod drives, rod run-in system, safety system and Servo amplifler '
( . system. l (4) One circuit provides power for the operation of the- f containment ventilation system isolation doors,' emergency . compres :or, truck entry door, and personnel airlock doors. 2.3 Surveillance Tests of Emergency Generator t The EG and the Cummins diesel engine are tested routinely
~
on the following basis: f (1) At least once a week the Cummins diesel engine which power.s the generator is started and allowed to run for a period of thirty minutes without load. L ~ l5 (2) In addition, the Cummins diesel engine is started and %- run for thirty minutes prior to each reactor startup following a shutdown greater than twenty-four hours, i (3) The ability of the emergency electrical generator to
; assume the emergency load is verifled on at least a
~-
semi-annual basis. Commerciat power to the reactor facility is interrupted at the transfer switch, simulating a complete loss of commercial power to the reactor L facility. This requires the EG to automatically start and assume full emergency electrical load. (4) The entire unit is serviced routinely as part of a planned
'l' preventive maintenance program.
LI L LI j III-16 f
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!2.4? Accident Analvata 2.4.1 Loss of Commerr ini Power with the Reactor Oneratina at -
i10 MW and the Rmeraency Generator F211a to Start Each system that is affected by a complete loss of electrical
; power is listed and commented upon in the following
~
paragraphs.
! (1) Reactor Control System At the time ofloss of commercial power while operating at 10 MW, the reactor would scram as a result ofloss of j
~. power to the electromagnets holding the blades in position. The blades would drop into the core by giavitational force and the reactor would be shut down.
(2) Reactor Process System ! All process systems (e.g., primary cooling, pool cooling, j etc.) will be placed in the shutdown condition to the yg. failsafe design of these systems. Ioss of electrical power g would cause a cessation of coolant flow cnd a closing of l- the isolation valves. In the primary system, redundant p I valves 546A and B open by spring actuation placing the in-pool heat exchanger in service. The fallsafe design of L the system permits shutdown decay heat removal vrith no
)h electrical power (reference: Appendix D of Addendum 4 i to Hazards Summary Report).
j (3) Containment Building Ventilation Isolation Doors . Power would be lost to the motor operated doors and. they would fail to close (or open) in response to any
- g; electrical signal. Also, the gasket seals would not inflate 3: since the inflating mechanism responds only when the-doors are closed against their stops. The backup cg isolation doors, however, fail closed upon loss of solenoid L
power and hence would automatically close upon loss of building power. g III-17
- l '
, (4) Emergency Air Comnressor -
, l
, g)l 'Ihe emergency compressor motor would fall to operate
'3 in response to a falling pressure in the reserve tank. 'lhe- '
reserve tank holds a volume of 10.5 ft3at a nominal'~
"(;
~
-pressure of 100 psi, which would be sufficient to inflate all gasket seals on all isolation doors if this were '
"; g.- ,
required. Bct the ability to recharge the tank to nominal E operating pressure would be lost in the event of a , complete' power failure. The primary function of this compressor is to provide air to the seal gaskets of all
'l"
. (-
isolation doors. Since the doors would not be operable
- g. with no. power there is little demand for the emergency ,
E, air supply. (5) Truck Entry Door: Door 101 Benmhole Floor During reactor operation and during periods when the ] ll l-reactor is left unattended, the door is closed and the seal is inflated.- Loss of commercial power would prewnt one . L from being able to open this door or deflate the' seal. r . IIence, loss of commercial power during normal reactor - operation would leave the status of this door unaffected. t l( u (6) Personnel Airlock Door: Doors 275/276 Grade Level' .! i During reactor operation and during periods when the:
- reactor is left unattended, one of these doors remains L ~
closed and the gasket inflated. Loss of commercial . power without the ability of the EG to provide emergency: power, would prevent one from operating these doors qk electrically. There is in existence, however, a procedure . i by which the gaskets can be deflated manually and the L doors manually opened or closed. Even though the doors L , cannot be operated electrically, it is possible for one to leave the building through these doors in the event of
- j. . ;
~
a power failure. However, the ability to maintain at least 4 one door in the closed position with its seal gaskets ,g inflated is lost if power to the doors is not available. The .E ^ containment integrity of the building, therefore, cannot j III-18 _ _ _ _ = _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ _ _ - -
I Il , be' guaranteed if emergency ~ and electrical power were not available to operate both doors.' Howevet, the reactor g will be shutdown and containment is not a vital lB' requirement. ) , (7) imhorntory and Reactor Exhaust Fana: ' EF-13 and EF-14. Fifth Level West Tower h Upon isolation of the reactor building, the operation or inoperation of these fans would have no consequence on the status of the reactor building. Upon~1oss of commercial building power without the availability of
, emergency power, both of these fans would cease to (l function.
(8) Reactor and Imboratory Corridor and Exit Liahts
. The 120-volt corridor and exit lights in both the reactor l' building and laboratory building depend upon commer-cial power or emergency backup. In most areas, emer-
.g; gency wall battery pack lights which operate upon loss of
;gL commercial building power provide sufficient lighting for
. all personnel to leave the reactor building and laboratory g corridor areas safely. Visibility in all the strategic place-ment of these battery pack emergency lights.
;(9) Fan Failure Alarm System
;g The exhaust fans, EF-13 and EF-14 " failure to operate"
>m; alarm system would not function. Loss of building commercial power and loss of emergency backup power
.l. would prevent the operation of these fans as previously discussed. Such a situation precludes the need for these
. alarms.
(10) Intercommunication System y I The laboratory area and the reactor building are provided with a muluple station intercommunication system. The
.; loss of this system results in the inability to transmit
.l III-19
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-j l -messages rapidly to the entire facility. However, .
telephone communication to each laboratory area and- !
- g: . various areas inside the reactor building would not be . ;
2Ei interrupted by a loss of commercial power, There is also: provided portable battery-powered transmitter-receiver - l 4
. packs which can be 2 sed to maintain communication i
' ~
between the Emergency Director in the laboratory loaby j and investigation parties sent out from that area. f
] t 4
(11) Reactor Bhilding Isolation and General Evacuation Marmi i, i Loss of reactor building power without emergency back-up power would result in the loss of all audible and visual t L( evacuation and isolation alarms. (12) Diesel Room Distribution Panel Power to the EG control panel, EG room lighting and EG : room temperature controls are provided by this panel.. Loss of commercial power as well an emergency power ' would leave these loads de-energted. However, failure of the generator engine to operp.te preempts the need for l these loads. l; (13) Ecactor Centrols and Instrumentation Power to all reactor instrumentation and controls, both 1
'l process and nuclev, is provided through a 120 VAC
, 7 distribution panel located in the control room. Loss of comn.ercial power without emergency backup would not effect the control and instrumentation power for 20 -
minutes because of the capacity of the uninterruptable I power supply (UPS). Once the: reactor is confirmed to be
~
shutdown and before the UPS batteries reach a low l voltage. condition, the control and instrumentation systems would be secured. The UPS unit would be i secured prior to the batteries reaching a low voltage condition to prevent a low voltage transient on the system. The reactor operators would then have no control console information relating to changes in the I III-20
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. - --._ .._._ 4 .. . . . . _ . . .
7-
' status of reactor system different than that gathered *
- when the UPS was operating.
' All' subsequent information regarding valve position and
- the status of the reactor would have to be obtained visually by the reactor operator. With the reactor shut down and with system status known before securing UPS power, the reactor operator can monitor valve position
;gl B indications at the reactor bridge and the redaced intensity'of the Cerenkov glow in the region of the core.
Operation of the emergency pool fill system would be unaffected. 2.4.2 Loss of Electrical Power during Reactor Shutdown Periods The status of all systems would be identical to that discussed in paragraph 2.4.1 with the following exceptions: (1) Reactor Control System In this case it can be assumed that the reactor is in the shutdown mode with all systems secured. This would be
. assured by the fact that prior to the loss of commercial
- (' power a complete shutdown checksheet for the reactor-and systems had been completed. Therefore, there would be no need for the reactor operator to determine 5 : the status of the reactor or reactor systems after the loss.
of commercial power. (2) Persons within Containment at the Time of Loss of
; Commercial Pmy,gr Research and other non-reactor staff personnel may be
=
within the containment building at tk a time of the loss of commercial power. All personnel allowed un-escorted access to containment have a knowledge of how to
- l. operate the personnel airlock door manually without assistance at a time when electrical power to these doors g is unaval'able. Simple directions are posted next to the u airlock doors.
- I 111- 2 1 1
I (3) Containment of the Ranctor Buildina Containment integrity of the reactor building would be l assured by virtue of the status of the reactor during a normal shutdown period. Truck entry door 101 on the beamhole floor would be closed and sealed, the 16" building exhaust isolauon valves would have failed closed, the supply and return fan on the fifth level would have
~l' ~
stopped operating, the primary isoladon doors would remain open, however, the backup isolation doors would have failed closed. 2,5 Conclualons ! Under the postulated failures, the reactor will shutdown and l the core will be cooled indefinitely by natural convection circulation through the in-pool heat exchanger. It may be necessar" ::o violate containment briefly to allow personnel to ! enter and exit containment, however, the reactor is in a safe configuration. Health Physics monitoring with portable instruraents would preclude the accidental exposure of l personnel to radiation. I 'I o lI i I I 1
- l I '
I, 111- 2 2 .
I l SECTION IV I PLANT AND SYSTEM MODIFICA'110NS l 1 July 1989 through 30 June 1990 i ModiScation 89 3: Containment Buildina and Air Saranlina System his modification provides a means for sampling containment i building air from outside the containment building during a Reactor I Isolation. 'Ihe need for a means of remote sampling of the containment j building (as opposed to having personnel enter into containment to sample) i I was recommended during a routine safety inspection by the Nuclear Regulatory Commission (NRC) on March 6-10, 1989. l
;g This modification utilizes the system originally installed to perform the reference volume method of determining containment buildmg leak rate. This system had not been used at MURR since 1978, w:.ien MURR l changed its method vf performing the building leak rate to the make up flow method. The reference cans for the reference volume system have I been disconnected to provide sampling points for the air sampling system (see Figures 1 & 2).
'Ihe safety evaluation for this modification documents that it does not present an unreviewed safety question as per 10 CFR 50.59.
Reference cans Containment inside containment sensing line Ne Valves p Isolation Valve [ x x 1 Hose connection i _ .. c__ u y l To Manometers IV-1 I -_
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IV-2
I ' I ModlScation 88-6: Emergency Electrical Power Unnrade (replacing ! cmergency generator (EG) and automatic transfer switch) : This modification replaces the original 45 KW gasoline powered EG with a 275 KW, diesel powered generator. This change was made due to the I age of the original EG as well as its lack of capacity for anticipated facility modifications, he HSR sections changed by this modification are mainly descriptions of the EG and the Emergency Electrical Distribudon system. ! ne HSR section (Addendum 5, Section 2.0) that contains an analysis of a complete loss of electrical power at MURR, remains virtually unchanged ! except for descriptions of the new EG and Emergency Distribudon System. t ne surveillance tests for the new EG will be incorporated into the HSR and ' will meet or exceed technical specification requirements. The HSR Addendum 5, Section 2.0 analysis concludes that a complete loss of I electrical power creates no hazard to the health or welfare of the general public. We bases for Technical Specification 3.10 states that on loss of normal electrical power the EG is not required for protection of the fuel t l element integrity. ,g The probability of the occurrence of a malfuncuon in the new EG is 3 expected to be less than that of the original EG, which had provided 22 years of service, but had no capacity for future expansion of emergency , electrical loads (i.e. new exhaust ventilation fans). : The safety evaluation for this modificadon, summarized here, I h documents that it does not present an unreviewed safety question as per 10 CFR 50.59. Modi 8 cation 88-11: Reactor Control Power Unnrade t This modificadon replaces the Elgar line conditioner with an ' Uninterruptible Power Supply (UPS) to provide regulated reactor control power. The replaced Elgar line conditioner was not specified in the original I- Hazard Summary Report. The line conditioner was added as an enhancement to the 120 VAC electrical supply to the reactor control and instrumentation power supplies. l lI l IV-3 I
1 HSR Addendum 1, Section 3.5 addresses the reguladon required of j MURR instmmentation and control and specifies system tolerances to l electrical supply changes over the following ranges' i Supply Voltage: 115 Volt +/- 10% Supply Frequency: 60 Hz +/- 5% l Temperature: 32'F to 120 F ,g These regulation parameters provided the impetus to install the Elgar line E conditioner, which can regulate supply electrical power as follows: .! I Supply Voltage: Supply Frequency: 115 Volt +/ .05% 60 Hz +/ .05% ne UPS can provide regulation as follows:
+/- 1% of Nominal Voltage (115 VAC) ,
+/- 0.1% of Nominal Frequency (60 Hz) ne UPS cannot provide the absolute voltage reguladon of a line conditioner, but it provides regulation over a greater range ofinput voltages.
A line conditioner regulates small monitoring voltage fluctuations, whereas a UPS regulates the outrut for even a complete loss of input voltage. , The UPS will protect the reactor instrumentation and control power
- from line transients, line noises and during the transitional period between the loss of facility electrical power and when the Emergency Generator I assumes emergency loads (see flg 3). The safety evaluation for this modification, summarized here, documents that it does not present an unreviewed safety question as per 10 CFR 50.59.
Modification 88-7: Exhaust Ventilation Uoprade l This modification to the ventilation system replaces the original exhaust fans with two speed fans and 100 H.P. motors and adds an acid I scrubbing system for the silicon laboratory to remove nitric, hydrofluoric and acetic acids. The new exhaust fans increase discharge flow from 20,500 SCFM to 33,500 SCFM, which ensures all MURR fume hoods will have a l minimum of 125 LFM face velocity. These modifications also required the addition of two (2) new roof top air handlers (6,000 SCFM cach) to supply air to the facility. Modificadons to the fan failure and warning system have I also been accomplished. l IV-4 I
I The fast speed isokinetic probes for the stack monitor were installed on May 7,1990. The ventilation system was balanced on May 14,1990 and placed in continuous fast speed operadon on May 15,1990. I 'Ihis modification will be complete when the auxiliary heating system for the new roof top air handlers are modified and operationally checked. The safety evaluation for this modification documents that it does not present an unreviewed safety question as per 10 CFR 50.59. I I I LI I I I i I o I IV-5 L
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I ! SECTION V I NEW TES13 AND EXPERIMENTS 1 July 1989 through 30 June 1990 i New expedmental programs during this period are as follows: I RUR 274 Experimenter: S.Gunn
Description:
This RUR authorizes die irradiation of Potassium l Hexachlorotridate (K2 IrCls). For the production of Ir 192 to be l used as a radioisotope tracer in oil exploradon to determine fracture patterns in rocks. I RUR 274A Experimenter: S.Gunn l
Description:
This addendum was added to allow the irradiation of Potassium Hexachlorotridate Iridium doped A1,p3, for the l production of Ir192. To be used as a radioisotope tracer in oil I exploration to determine fracture patterns in rocks. RUR 275 Experimenter: S.Gunn l
Description:
his RUR authorizes the production of S35, by the irradiation of Potassium Chloride. This isotope is used to label i organic compounds for industry. l RUR 277 Experimenter: S.Gunn ,l
Description:
This RUR authorizes the irradiation of StronUum l Carbonate to produce Sras and Sr89. These isotopes are used by radiochemical companies in blood flow studies. I RUR 278 Experimenter: S.Gunn
Description:
This RUR authorizes the irradiation of Lutetium l Oxide enriched to in Lu17e up to 100%. Used for research testing and labeling anubodies for cancer studies, pI-RUR 282 Experimenter: S.Gunn ,
Description:
This RUR authorizes the irradiation of Osmium metal, natural and enriched. The Osmium is used to produce l Ir191m for heart studies in implants. l V-1 I
l RUR 285 Experimenter: S.Gunn Descripuon: his authorizes the irradiation of a thin UO2 i (Uranium Oxide) deposition on utanium foil. This will be used
- for Nuclepore's new irradiator case and are test runs. ,
e I RUR 286 Experimenter: S.Gunn Descripuon: This authorizes the irradiation of Antimony Oxide (Sb2 0 )4 for industrial use as a tracer in oil exploration to determine fracture patterns in rocks. '
.i I
RUR 287 Experimenter: S.Gunn
Description:
This authortnes the irradiation of Scandium ; Chloride (ScCl3 ) for the producuon of Sc46. his is used as a industrial tracer for oil exploration to determine fracture patterns in rocks. '
' Each of these experiments has a written safety evaluation on flie which provides the basis for the determinadon that it does not involve an -
unreviewed safety quesdon as per 10 CFR 50.59. I - iI LI I I I LI LI V-2 I
. .. . -- _ 1 I ;
'I SECTION VI SPECIAL NUCLEAR MATERIAL ACTIVITIES 1 July 1980 through 30 June 1990 i
- 1. SNM Receipts: A total of 22 new fuel elements were received from Babcock and Wilcox (B & W), Lynchburg, Virginia.
I Shipper Elements Grams U Grams U-235 '
. B&W MO241, MO292, MO293, MO294, 18,256 17,008 '
MO295, MO296, MO297, MO298, MO299, MO300, MO301, MO302, i MO303, MO304, MO305, MO306, MO307, MO308, MO309, MO310, MO311 MO312
- 2. No Spent fuel elements were shipped.
- 3. Inspections: 'Ihere were no Physical Security and Special Nuclear
- l Material accountability inspections conducted by the Nuclear Regulatory Commission.
- 4. SNM Inventory: As of 30 June 1990, MURR was financially responsible for the following DOE owned amounts:
I Total U = 49,788 grams Total U-235 = 44,246 grams included in these totals are 36 grams of U and 34 grams of U-235 non- /' I fuel, DOE 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. ,f-I I. m I
*t. ,w-. ,4.#e.
'I I 'Ihe fuel elements on hand have accumulated the following burnups as of 30 June 1990: Burned-un Elements Element No. MWD Element No. M Element No. hG@, MO-232 145.46 MO-260 134.19 MO-274 148.75 I MO-234 MO-239 145.46 145.05 MO-261 MO-262 135.40 129.63 MO-275 MO-276 144.91 148.04 MO-240 145.05 MO-263 135.36 MO-277 146.73 MO-246 143.80 MO-264 129.63 MO-278 148.04 MO-247 146.78 MO-265 135.36 MO-279 146.73 I MO-248 MO-249 143.80 146.78 MO-266 MO-267 132.01 123.08 MO-280 MO-281 146.34 143.00 MO-254 145.15 MO-268 132.01 MO-282 146.34 l MO-255 MO-256 126.70 146.15 MO-269 MO-270 123.08 144.50 MO 283 MO-284 143.00 148.85 MO-257 126.70 MO-271 144.50 MO-285 I MO-258 MO-259 134.19 135.40 MO-272 MO-273 148.75 144.91 MO-286 MO-287 146.76 148.85 146.76 Elements in Service MO-241 119.19 MO-297 72.59 MO-307 34.22 I- MO-288 112.41 MO-298 103.68 MO-308 15.64 MO 289 140.88 MO-299 67.45 MO-309 34.22 I MO-290 MO-291 112.41 140.88 MO-300 MO-301 50.98 67.45 MO-310 MO-311 15.64 0,00 MO 292 81.72 MO-302 50.98 MO-312 0.00 MO-293 119.19 MO-303 38.73 MO-313 0.00 MO 294 81.72 MO-304 39.45 MO-314 0.00 I MO-295 MO-296 72.59 103.68 MO-305 MO-306 38.73 39.45
-l Average Burnup: 109.67 MWD I
I I VI-2 I
"I SECTION VII REACTOR PHYSICS ACTIVITIES ! 1 July 1989 through 30 June 1990 l
\
- 1. Fuel Utilization: During the period,1 July 1989 through 30 June 1990, the following elements reached licensed or feasible burnup and were retired:
Serial Number Final Core Date fut Used MWD l MO258 MO260 89-42 89-42 08-14-89 08-14-89 149.49 149.49 l 1 g MO262 89-56 10-23-89 149.80 3 MO264 89-56 10-23-89 149.80 : MO266 89-41 08-07-89 149.81 MO267 89-48 09-09-89 146.32 , MO268 89-41 '08-07-89 149.81 MO269 89-48 09-09-89 146.32 MO270 89-48 09-09-89 144.50 ~ MO271 89-48 09-09-89 144.50 I MO272 MO273 90-20 90-02 04-30-10 01-15-90 148,75 144.91 MO274 90-20 04-30-90 148.75 ,l MO275 90-02 01-15-90 144.91 MO276 89-66 12-26-89 148.04 MO277 89-64 12-11 89 146.73 MO278 89-66 12-26-89 148.04 MO279 89-64 12-11-89 146.73 MO280 90-13 03-19-90 146.34 MO281 90-19 04-23-90 143.00 MO282 90-13 03-19 90 I MO283 MO284 90-19 90-21 04-23-90 05 07-90 146.34 143.00 148,85 J - MO285 90-27 06 11-90 146.76 MO286 90-21 05-07-90 148.85 MO287 90-27 06-11-90 146.76 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 Vll-1 I '
=
I i elements of which only the inidal critical fuel elenient serial numbers are listed in the following table. To increase operating efficiency, fuel ; l- elements are used in mixed' core loadings. nierefore, a fuel element ! , fabrication core number is different from its core load number. : i . iI Fabrication Core No. 52 Serial No. MO241 InitialCore Load No. 89-41 Initial Ooeratina Date 07-31-89
- 52 MO292 89-45 08-28-89 !
52 MO293 89-41 07-31-89 > 52 MO294 89 08-28-89 53 MO295 89-51 09-18 89 53 MO296 89-54 10-10-89 53 MO297 89-51 09-18-89 f
.g 53 MO298 89-54 10-10-89 ,
53 MO299 89-61 11-20-89 - 53 MO300 90-02 01-08-89 53 MO301 89-61 11 20-89 I 53 54 54 MO302 MO303 90-02 90-10 01-08-90 02-19 90 MO304 90-13 03-12-90 54 MO305 90-10 02-19-90 54 MO306 90-13 03-12-90 g 54 MO307 90-20 04-23-90
- g. 54 MO308 90-24 05-20-90 54 MO309 90 20 04-23 90 !
54 MO310 90 24 05-20-90
- 2. Fuel Shipping: No spent fuel was shipped.
- 3. Fuel procurement: Babcock and Wilcox. Lynchburg. Virginia, is MURR's fuel assembly fabricator. This work is contracted with the U.S.
Department of Energy and administered by EO&G Idaho Inc.. Idaho Falls. Idaho. As of 30 June 1990,115 fuel assemblies fabricated by Babcock i f and Wilcox had been received and 111 used in cores at 10 MW. < LI LI VII-2
-,r -
*- '~
9 ---7 -v
I ..'l - 4. Licensing Activities:- On 6 June 1990, the Nuclear Regulatory Commission approved Amendment No.19 to the Facility Operating License No. R-103. This amendment temporarily increased the Special Nuclear Material I hwentory under Facility license R-103 pending the establishment of capability for the offsite shipment of spent fuel. As part of its annual reporting requirements, the status of establishing this spent fuel l shipping capability is described below: On April 18,1990, MURR submitted a request to the Nuclear Regulatory
.I Commission (NRC) to amend the BMI-1 spent fuel shipping package Certificate of Compliance to authorize MURR to use this cask. On July 6 1990, the NRC submitted four questions to MURR concerning our I- request. The answers to these questions are being prepared and will be sent to the NRC in mid August.
I A submittal was made September 12, 1986 pertaining to the new MURR I fuel design with associated revisions to the Technical Specifications. The fuel design has been evaluated by the NRC and Amendment No. 20 was issued August 1,1990 approving the new fuel design. A description l of this amendment will be provided in the 1990-1991 annual report. A request for a unique purpose exemption as defined in 10 CFR 50.2 I was submitted September 26,1986 and is pending.
- 5. Reactor Characteristic Measurements:
Sixty-five refueling evolutions were completed. An excess reactivity verification was performed for each refueling and the average excess reactivity was 1.88%. The largest excess reactivity was 2.64%. MURR I Technical Specification 3.1(f) requires that the excess reactivity be less than 9.8%. l ReactMty measurements were performed for 10 evolutions to verify reactMty parameters for the flux trap. The complete worth curves of the I four shim and one regulating blade were obtained after the beryllium reflector changeout in November,1989. I Vll 3 r I
I Physical inspecuons of the following fuel elements were perfonned to verify operational parameters. MO 262 from Core 44 8/11/89 MO-264 from Core 49 8/11/89 MO-283 from Core 51 11/15/89 MO-280 from Core 51 2/8/90 MO-272 from Core 50 4/18/90 . All measurements were within operadonal requirements. I I I I
- I I
I I I
.I I
LI vu.4 I
- I I SECTION VIII I
SUMMARY
OF RADIOACTIVE EFFLUENT RELEASED 'IV THE ENVIRONMENT I Sanitary Sewer Efiluent 1 July 1989 through 30 June 1990 l Descending Order of Activity Released for Isotope Totals > 1.00E-5C1 g Nuclide Amount (Cil Nuclide Amount (Cil H-3 5.55E 01 W- 188 1.23E-04
.I S-35 1.47E-02 Rh-105 1.13E-04 Zn-65 7.10E 03 Eu-154 8.90E 05 As-77 4.07E-03 Fe 59 8.07E-05 Ca-45 3.40E-03 Nb 95 8.01E-05
- Co-60 1.24E-03 Pt-191 7.26E-05 Cu-64 9.82E-04 Cu-67 6.32E-05 Tc-99m 9.81E-04 Rb-86 3.93E-05 Re-186 9.54E-04 K-42 3.92E-05 Sc-46 9.52E-04 Zr-95 3.36E 05 Cr-51 7.25E 04 Sb-124 3.25E-05 Sm-153 6.36E-04 Ho-166 2.65E-05 Ag-110m 4.61E-04 Mn-54 2.54E-05 l
Eu-152 3.86E-04 Ta-182 1.96E-05 I Ba 131 Se-75 3.37E-04 2.62E-04 Sb-122 Pa-233 1.81E-05 1.28E-05 Na-24 2.39E-04 Gd-159 1.27E-05 I Cd-109 2.23E-04 I I Vill-1 I
I I l I Stack Efiluent 1 i 1 July 1989 through 30 June 1990 Ordered by % Technical Specification (TS) Limit ! I Isotope Tot. Release FY 89-90 IC1) Average Concentration fuCI/ml)
'IS Limit (x MPC)
% TS*
I Ar-41 I-131 K-40 5.9E+02 5.4E-04 1.3E-04 2.0E-06 2.1E 12 5.1E-13 350 1 14.286 2.121 ' 1 0.513 ; I Eu-154 Os-191 Eu-152 3.7E-05 2.8E-03 4.9E-05 1.5E-13 1.1E-11 1.9E-13 1 1 1 0.145 0.110 0.048 W-188 9.9E-06 3.9E-14 !I I-135 H-3 1.2E-02 2.3E+00 4.8E-11 8.8E 09 350 350 1 0.039 0.014 0.013 Co-60 9.2E 06 3.6E-14 l I-133 Se-75 4.1E-03 9.1E-05 1.6E-11 3.6E 13 350 1 1 0.012 0.012 0.009 I-134 4.6E 02 1.8E- 10 350 0.009 i V-52 2.3E 01 8.9E- 10 350 0.008 l
.l Cd-109 2.5E-05 9.8E-14 1 0.005 l
Ce-139 1.1E-06 4.5E 15 0.004 ; I 1 Pd-103 3.2E-04 1.3E-12 1 0.004 ' Te 125m 3.7E-05 1.4E- 13 1 0.004 I-132 8.9E 03 3.5E- 11 350 0.003 Cl38 1.8E-01 6.9E- 10 350 0.003
-I r
Cs-137 3.5E-06 1.4E 14 1 0.003 Ce 144 1.3E-06 5.0E-15 1 0.003 > 3 Hg 203 2.5E-05 9.7E 14 1 0.002 lJ 1 Co 57 Zr-95 2.9E-05 4.4E 06 1.1E 13 1 0.002 1.7E- 14 1 0.002 , Ba-140 4.3E-06 l Xe-135m Eu 155 3.7E-02 9.0E-06 1.7E- 14 1.5E- 10 3.6E-14 350 1 1 0.002 0.001 0.001 Sc-46 2.1E 06 8.2E 15 0.001 lj- Te- 123m Ta-182 1.7E-07 1.6E-06 6.8E- 16 6.2E-15 1 1 0.001 1 0.001 Total 17.381
- Isotopes observed at <0.001% TS limit not listed.
1
~
Stack flow rate - 20,500 ft.3/ min, from 7/1/89 to 5/13/90. Following ventilation upgrade, stack flow rate - 33.500 ft.3/ min. from 5/14/90 to g 6/30/90. (See Section IV. Plant and System Modifications) I ""- t
I Amended Table
- Stack Effluent !
l 1 July 1988 through 30 June 1989 Ordered by %'IS ! I Isotope Tot. Release FY 88-89 Average Concentration TS Limit % TS* i' (C11 fuC1/ml) fx MPC) Ar-41 9.2E+02 3.6E-6 350 25.667 I-131 6.1E-4 2.4E-12 1 2.417 ! l Eu 154 H-3 1.0E-5 2.8E+0 3.9E-14 1.1E-8 1 350 0.039 0.017 , I-135 1.0E-2 4.0E-11 350 0.011 ,l I-133 4.0E 3 1.6E-11 350 0.011 l Co-60 7.5E-6 3.0E-14 1 0.010 Eu-152 9.5E-6 3.7E- 14 0.009 l 1 Cd-109 2.9E-5 1.1E-13 1 0.006 I-134 2.9E-2 1.1E-10 350 0.005 Ce-144 2.GE 6 1.0E-14 0.005 l l 1 Ce 139 1.2E-6 4.7E- 15 1 0.005 Se-75 4.5E-5 1.8E-13 1 0.004 V 52 1.1E 1 4.3E-10 350 0.004 !l l I-132 Hg-203 7.6E 3 2.5E-5 3.0E-11 1.0E-13 350 1 0.003 0.003 - L Na-22 1.8E 6 7.0E-15 1 0.002 Cs-137 2.0E-6 8.0E-15 1 0.002 ' Te-125m 1.2E-5 4.7E- 14 1 0.001 Xe-135m 2.3E-2 9.2E-11 350 0.001 Br-82 4.6E-3 1.8E-11 350 0.001
- Zn-65 2.8E-6 1.1E-14 1 0.001 Other isotopes observed at <0.001% TS limit.
Stack flow rate ~ 17.100 cfm. An error in data entry for the assembly of the 1988/89 Stack Effluent Table was discovered following submittal of the 1988/89 Operations Annual Report. The error involved adding the same isotope data more than one time which inflated some of the % TS totals. Several of the corrected totals were < 0.001% and so some of the isotopes originally reported are not listed in this corrected table. I ; l Vill-3 I
I ; SECTION IX
SUMMARY
OF ENVIRONMENTAL SURVEYS 1 July 1989 through 30 June 1990 I Environmental samples are collected two times per year at eight locations (HP-11: " Environmental Sampling") and analyzed for radioacuvity.
'Ihese locations are shown in Figure 1. Soll and vegetation samples are I taken at each location. Water samples are taken at three of the eight locations. Results of the samples are shown in the following tables.
g 1. Sampled during October 1989. Detection Limits * - Matrix Alpha Beta Gamma Tritium Water I Soil 1.0 pC1/1 1.0 pC1/g 2.8 pC1/1 4.5 pC1/g 219.0 pCi/l 1.9 pC1/g 19.3 pC1/ml N/A of sample Vegetation 2.0 pCl/g 9.3 pC1/g 3.4 pC1/g 19.3 pCi/g ' distillate
- Gamma and tritium analyses are based on wet weights l while alpha and beta analysis are based on dry weights.
Determined Radioactivity Izvels Vegetation Samples Sample Alpha Beta Gamma Tritium (pct /g) fpCt /g) (pci/g) (pC1/g) 1 -V-36 < 2.0 26.5 < 3.4 < 19.3 2-V-36 < 2.0 19.5 < 3.4 < 19.3 i l 3-V-36 < 2.0 28.5 < 3.4 < 19.3 4-V-36 < 2.0 12.8 < 3.4 < 19.3 l 5-V-36 < 2.0 23.6 < 3.4 < 19.3 L 6 V 36 < 2.0 < 9.3 < 3.4 < 19.3 i 7-V-36 < 2.0 15.7 < 3.4 < 19.3 10-V-36 < 2.0 23.0 < 3.4 < 19.3 L l IX-1 I
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I l Determined Radinactivity Izvels ! Soil Samples ; Sample Alpha Beta Gamma ' foci /g) foci /g) foci /g) 1-S-36 < 1.0 21.0 5.4 2-S-36 < 1.0 11.7 6.1 ; 3-S-36 < 1.0 9.3 4.2 4-S-36 < 1.0 7.9 4.1 ! 5 S 36 < 1.0 13.2 6.6 ; 6-S-36 < 1.0 < 4.5 < 1.9 > 7-S 36 < 1.0 10.8 5.5 10-S-36 < 1.0 10.7 5.8 ; I Determined Rndinactivity IEvels Water hmples .E Sample /.'pha Beta Gamma Tritium E foC1/l) foC1/11 foC1/11 ,,foC1/ml) 4-W-36 < 1.0 < 2.8 < 219.00 < 19.3 6-W 36 < 1.0 < 2.8 < 219.00 < 19.3 , 10 W-36 < 1.0 7.S < 219.00 < 19.3
- 2. Sampled during May,1990 Detection Limits * '
Matrix Alohn Beta Gamma Tritium Water 1.0 pC1/1 4.0 pC1/1 228.5 pC1/1 22.5 pC1/ml 'I Soll 1.0 pC1/g 4.0 pC1/g 1.5 pCi/g N/A of sample
' Vegetation 1.9 pC1/g 8.0 pC1/g 42 pC1/g 22.5 pC1/g !
I' distillate
~*
Gamma and tritium analyses are based on wet weights I while alpha and beta analysis are based on dry weights. . h IX-3 I
I Determined Radioactivity Izvels Vegetation Samples I Sample Npha foci /g) Beta foci /g) Gamma foci /g) Tritium fact /a) 1-V-37 < 1.9 35.6 < 4.2 < 22.5 2-V-37 < 1.9 28.8 < 4.2 < 22.5 l 3-V-37 < 1.9 18.2 < 4.2 < 22.5 4-V-37 < 1.9 36.1 < 4.2 < 22.5 I 5-V-37 < 1.9 29.7 < 4.2 < 22.5 6-V-37 < 1.9 20.4 < 4.2 < 22.5 7-V-37 < 1.9 26.0 < 4.2 < 22.5 10-V-37 < 1.9 29.7 < 4.2 < 22.5 I Determined Rsidloactivity Izvels soil samples Sample Mpha I (pct /c) Beta foci /n) Gamma foci /nt, 1-S-37 < 1.0 19.6 7.6 I 2-S 37 < l.0 22.6 8.9 3 S 37 < 1.0 18.9 5.0 4-S-37 < 1.0 10.0 4.9 5-S-37 < 1.0 19.5 7.9 6 S-37 < 1.0 10.6 4.2 7-S 37 < 1.0 15.9 6.9 10-S-37 < 1.0 19.9 8.2 I I: l IX-4 g 1
I Determined Radioactivity Izvels Water hmples
- Sample Alpha Beta Gamma Tritium (p01/11 ,fpci/11 (p01/11 fpC1/ml) 4-W-37 < 1.0 < 4.0 < 228.5 < 22.5
~
6 W-37 < 1.0 6.2 < 228.5 < 22.5 10-W-37 < 1.0 9.8 < 228.5 < 22.5 ENVIRONMErfrAL TLDs An environmental TLD program was initiated in April 1990 to monitor environmental dose rates surroundlag MURR. These TLD monitors are being changed on a quarterly basis. Results from these measurements will be reported in future summaries of environmental surveys. NUMBER OF FACILrlY RADIATION AND COffrAMINATION SURVEYS Radiation Surface Contamination Air Samples ISM
- I- July 54 46 21 August 42 34 23
- September 46 37 21 October 34 29 22 I November December 36 36 30 33 22 21 M
January 31 25 23 i February 29 27 20 March 47 44 22
'g April 40 32 21 23 May 44 48 23 M M June 2.1
-l TOTALS 475 421 260
;g 96 Radiation Work Permits were issued during the fiscal year 1m IX-5 1
I !l Miscellaneous Items-In December 1989, Mr. Ronald J. Dobey, Jr., Health Physics I Technician, completed his MBA degree and left employraent with the University to become an Agreement State Regulator in Kansas. Mr. R. I Mark Stumbaugh was hired as a Health Physics Technician in March 1990. Mr. Stumbaugh received his original training in the Nuclear Navy Program, followed by two years in health physics jobs at several nuclear power l plants. An additional Health Physics Technician position waa established and hiring for that position is underway. Ms. leslie M. Powell was hired - I into the newly established secretarial position to support the MURR Health Physics Group. She had been the MURR receptionist for over three years and was familiar with many of the secretarial needs of the Health Physics l Gmup. Three new Health Physics Standard Operating Procedures were I added. HP-34: "MURR Identification Badge Program and Implementation" was established to provide guidelines and policies for the use and issue of I identification badges at the MURR facility. HP-35: "Beamport Area Access" was established to define access controls for the area surrounding the beamports following the redefinition of the MURR neutron beams as high l radiation areas (see Inspection Report No. 50-186/89001). HP-36:
" Remote Sampling of Containment Air" was established to provide the I capability of sampling containment building air without entry in the cases where airborne contamination is suspected following containment isolation. Copies of these procedures are attached at the end of this section.
ADCO Services, Inc. has continued to act as our institutional waste broker. Through ADCO, MURR disposed of 798.9 cubic feet of LSA material generated at MURR. The replacement of the reactor beryllium reflector and graphite reflector wedges (see Section I) was accomplished with a total of 4.96
.k manrem exposure. The October 1981 changeout of the heryllium reflector had required 14 manrem to complete the job (see Operations Report I 1981/82). The prejob review and training utilizing the knowledge gained from this 1981 job helped to identify several ALARA measures which were implemented for the 1989 job: this accomplished a 9 manrem reduction.
I m.e I
I l l Review of dose received for the 1989 job have identified additional ALARA measures to be considered for the next scheduled beryllium reflector I changeout. A program to improve the docurr.9ntah. . :,onnel dose review g . for the purpose of ALARA assessment ?tas b # . . Afarch 1990. 'Ihis program has established ALARA reportit.g criteria and provided an 3
-improved mechanism for management review. As the program is being l established and tested, assessment is continuing in how to best fit the ALARA review process for the various working groups at MURR. ALARA I efforts to reduce extremity exposures in the production of radioisotopes used in radiopharmaceuticals at MURR have continued. Changes in the chemical compound of the target materials have allowed for reduced handling of these materials during processing, which has greatly reduced 1 the extremity exposures received per acuvity handled.
I I I 1 I I I I-I iI 1 l IX-7
CHANGES 'ID HEALTH PHYSICS STANDARD OPERATING PROCEDURES r MADE FROM 1 JULY 1989 THROUGH 30 JUNE 1990 l 'Ihere were three new additions to the Health Physics SOP manual during the year (#HP-34, #HP-35 and #HP-36). These additions are i contained in this section. I, , i l I lI I-3 I . I IX-8 I .
I SOP MP-34 Rev. new Page 1 of 3 Appr'd b N. b - A Date R/1/R9O MURR Identification Badge Program and Implementation I. Policy All personnel with unescorted access to the facility shall wear I.D. badges to aid in maintaining the security of the MURR facility by preventing unauthorized access. II. Purnene To provide guidelines and policies for the use and issue of identification badges at the MURR facility. III. ceneral Personnel at the facility will be issued I.D. badges based on I level of access, blue for containment access and orange for lab access. Personnel having containment combination access will be identified by a special hole punch in the white side of the badge. This will assist in maintaining positive I control of the combination. Along the bottom of the badge the individual's name will be printed. Temporary badges will also be available for use as spares and I- for personnel who have unescorted status but visit the site infrequently. These badges will appear much like the 3 permanently assigned badges with the following exceptions: in g place of a photo the word " TEMP" will be visible; and instead of a name a serialized number will be used for identification at the bottom of the badge. These will be maintained at the 'I Front Desk and in the Control Room and will be logged out as needed. Personnel requiring an escort will be recognizable by their I' lack of an I.D. badge. All personnul at the facility without an I.D. badge shall be escorted by badged personnel. IV. Re seen s mility Inside the facility, I.D. badges are to be worn at all times l' and shall be worn visibly on the front of the body. visors will be responsible to ensure that people in their group wear their badges. Super-I I
r- , I I SOP MP *s a I Rev. new Page Appr,d g x . k A f 2 of 3 Date R/1/ I Lost, misplaced, or defaced badges,will be replaced at a cost of $2.00 (cost of materials) for the first replacement and
$5.00 each subsequent replacement. Chronic loss or I forgetting of badges will be reviewed and could result in loss of access to the facility.
During non-working hours each person will be responsible for the safe, secure keeping of the issued I.D. badge. Badges shall be kept in a manner which will preclude theft or misuse of this identification material. They should not be
.g hung from rear view mirrors or left on car seats or in any
'g unsecured area where theft could occur.
At the time of termination of access to the MUR3 facility, the individual and sponsor will ensure that the I.D. badge is turned in to the receptionist for proper disposition. ,3 Any unescorted personnel inside the facility without an I.D. '5 badge shall be challenged and the Control Room notified of the unauthorized entry. The individual shall then be i 3 escorted to the lobby for interrogation by the receptionist ! l-g or reactor operator. This applies as well to a person with lab access I.D. badge found unescorted inside containment. V. Procedure The materials used to produce the badges will be assigned as
~g follows: ID camera and film to Health Physics; and the badge g blanks, lamination folders, die cutter, and lamination . machine to the Director's Office. The name or number will be !
printed by Kroy lettering. This will be done with cooperation ) between HP and the Director's Office. j Issuance of identification badges will be as follows: the I individual will be photographed following Indoctrination and this photo will be given to Direct <-- . .ece personnel for subsequent badge assembly. At this time the new personnel will be given a temporary I.D. badge while awaiting permanent LI l I.D. badging. The new personnel should normally receive their permanent I.D. badge within twL weeks. l The designation of combination access will be made by the Reactor Manager. As such, the 'mol used to delineate this will be maintained in the Cont i Room. Once an individual lO is on the list of persons authurized to have containment ,g I
\
r 4 q ..c. r
,;T 1
[ SOP HP-34 E Rev. new Page 3 of 3 Appr'd 6 N.le4- * ! b Date~ R/1/e W r\ j E entry, the-operators will be able,to give the combination and pg make the designation symbol on the I.D. badge, 1: , igJ During normal working hours, issuance of temporary I.D. ' g: badges will be done at the Front Desk in the lobby. The t receptionist will verify the individuals access level and-then' log out by serial number a temporary badge to the person based on access. level. After normal. working hours temporary i badges shall be obtained by contacting the Control Room (#13)
.'The' operations personnel on shift will verify access level and issue the-appropriate I.D.' badge and log the badge out.to the individual. Temporary badges shall be returned at the l
end of the use to~the respective issuing place. E u l l 1 LI II
- I I
I; I ,
!I .
.a.-----_-_--_----.--.-.-----s-_-.-_.--,-_:.- _.- - - - _-- _ _ _ __ _ _ --- --- - - '* w
~
I ' SOP RP-35 Rev. new Page- 1 of 3 - Appr'd 6. M 2 Date 9/1/8O-g Beamport Area Ac, cess Li . I. Pn11cv Unescorted entry into the beamport area shall be limited to
.-. those individuals approved for access to the area. All other individuals entering the beamport area shall be escorted by
.=
-individuals with approved access.
II, Purname I 'The purpose of controlling access to the beamport-area is to ensure compliance with 10 CFR 20.203 (c) : "High radiation areas." III. n.finitions Beasport Area is defined as that area around=the biological
. shielding controlled by fencing-with key access.
Radiation Area means any area, accessible to personnel, in which there exists radiation at such levels that a major
.I portion of the body could receive in any one hour a dose-in excess of 5 mrem, or in any 5 consecutive days a dose in excess of-100 mrem.
Righ Radiation. Area means any area, accessible to personnel, in which there exists radiation at such levels that a major portion of the body could receive in any one
~I' hour a dose in excess of 100 mrem.
Authorized Individual means any individual who is
'I' identified as having need of access to the beamport area, and who has successfully completed training and been approved for I" IV. Responsibi1Itles A. Access Control
- 1. Limited to authorized individuals.
- 2. Guests approved by Health Physics prior to escorted entry by an authorized individual.
I .... .
e SOP HD-35 Rev. new Page 2 of 3 Appr'd I N.Y Date 9/i/aO
- 3. Access control required,during reactor operation.
- 4. List of authorized indivi' duals maintained at each access point.
-: 5 .- Authorized individuals responsible for ensuring positive control of each entry.
B. Training
. ,m ; 1. Initial request made to Health Physics by sponsor lgL to complete training of individual to be authorized.
- 2. Successful completion of training and-approval-for
..'$ individual to be authorized as per TRAINING 2:
5 " General'Information for Beamport Area Access - Radiation' Safety."' C. Door Control
- 1. During reactor operation, doors.shall remain locked except during periods when access to the area'is I.. required, with positive control over each entry.
- 2. .During reactor shutdown, doors may be opened to individuals with containment access unless.special-IL maintenance procedures warrant limited access.
- 3. Reactor Operations will ensure doors are closed prior to reactor startup.
D. Posting
- 1. Entry points to the beamport area shall be posted as per 10 CFR 20.203 (c) (1) with the phrase, "in ' the:
L. beam", added.
- 2. High radiation areas shall be limited to in the neutron beams during normal operation.
- 3. Neutron beams ac assible to personnel shall be i , conspicuously marked with yellow plastic streamers.
1 g-. _.p .
..,4 t
SOP MP-35 Rev. new Page 3 of 3 Appr 'd I. M . IM J '
^
Date 9/1/89'} ,
; E. Radiation Surveys ,
- 1. Beamport area shall be included in the monthly- . ;
survey of the beamport floor performed by Health Physics.
- 2. Copies of.'the monthly area surveys are posted on the beamport floor bulletin board and should be reviewed by the authorized individuals.
- a 3.. ' Survey instruments are available for authorized L
- g . individuals to make radiation measurements.
-4. Changes or movement of the. physical arrangement of 1 l .
the instruments, beam stops, and shielding'shall be' l
.5L done in cooperation with Health Physics to monitor-for changes in dose rates or contamination. L 5
- 5. Items- removed from the beamport area suspected to be activated and/or contaminated shall be monitored
- by Health-Physics prior to.being released or stored. ,
LI LI r I: . I. LI: L l 'I I!
- l
~
q y i , SOP MP-36 Rev, new Page 1 of 3 I'- Appr'd b M b L f Date 9/la Remote Sampling of Containment Air I: . 6 I. Pollev When, containment isolation has occurred and there is K indication that airborne contamination exists, an estimation. of..MPC shall be made before reentry of personnel into con-tainment, except for cases of life-saving and accident mitigating activities (SEP-ll). II. Purname This procedure will be used to obtain air sample (s) z$- using the remote containment sampling station. . The sample (s) W -- will then be analyzed to estimate MPC levels prior to reentry , of personnel into containment.
.III. Scona l
This procedure will apply in cases where airborne con-
- tamination is suspected following containment isolation.
Airborne contamination will be suspected by increased stack monitor readings (s), sustained increases in several of the s l[ ARMS readings, or by the circumstances leading to the isola-- 3- tion. This procedure will not be required in the cases of life-saving and accident mitigating activities-(SEP-11). Procedure IV, A. The remote sampling station is located in the basement area next to the pneumatic tube blower system. A . schematic for the remote sampling station is shown in !. Figu:e 1. B. Obtain an air pump, hose connectors, and tubing adequate to sample the system. Connect pump hose to sample from the reference can line (through Valves 1 and 2) and
- exhaust to the containment sensing line (through Valves 3 and 4).
C. Open isolation valves in number sequence and flush the system (sample line v.olume is approximately 0.5 f t3) . Monitor dose rate around pump during this flush. g -
, j . . _ . . _ _ _ . .._. _ _ _ _ _ _ _ .
3g u
.3: -i SOP MP-36' .;
-Rev. newPage 2 of 3 '
Appr'd d.M ^ r\ i- Date 9/is/s
.D. Close' valve 2, and with the pump still running, disconnect hose .t Valve 2. Connect evacuated air , sample can at vaAve 2 connection, and open Valve 2 and
'n.
-the sample can valve to obtain containment air sample.
Close both valves, disconnect sample can and reconnect pump-hose. If dose rates are acceptable,: Valve 2-may be ; c .
, opened and.the system allowed to continuously. flush in preparation for any further sampling.
l E. Analyze sample can via gamma spectroscopy to identify-isotopic contamination and estimate MPC levels in
' I -' containment air. -This information will be included.in the review'of precautions required for reentry into-containment or recovery from containment isolation.
F. This remote sampling station may also be utilized to
. obtain other types of air samples, such as filters. !
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ar Ii sop MP-36 Page' 3 Ef 3 I Rev.' new Appe'd S k l^- N u l :: I.< ; - Date~ 9/is/no [{- e . Reference cans L Containment l'(
.inside containment sensing line W 4 m
- U-
. >fC >80- ,
Isolation valves & ig a
. 7 x 1"*" ~ *" ;
LI: > lI b Hose Connection J Manometers ? Figure 1. Remote Sampling Station for Containment Air l L LI:. L_ .
' 1 I -
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~ ; M .: :M. E - .~ E c E ,E 1M - E WL :EI E~ -E :W> lW E _ E; E 'E!- E SECI1ON X - -i l
l
SUMMARY
OF RADIK110N EXPOSURES TO l FACllIlY STAFF. EXPERIMENIERS AND VISTTORS
.1 July 1989 through 30 June 1990 l
l 1. Largest single exposure and average exposure are expressed in millirem. ( 2. Minimal exposure is defined to be gamma < 10 mrem: beta < 40 mrem; neutron < 20 mrem.
- 3. ME = Number of monthly units reported with minimal exposure.
! 4. AME = Number of rrionthly units reported with exposure above minimal l
- 5. AE = Average mrem reported for all units above minimal.
! 6. IIE = liighest mrem reported for a single unit for the month. l l PERMANENT ISSUE FIIM-BADGES i Beta. Gamma. Neutron (Deco Dosel Whalehadv "Aa. l JML AUG SEE - QCE M DEC JAN EER MAR AER MAX M ME 120 125 107 104 116 129 115 109 131 136 129 109 j AME 48 53 64 69 58 45 61 70 53 45 63 61 AE 67 68 66 56 100 60 79 63 73 82 74 67 l IIE 260 210 280 170 400 190 270 390 320 250 270 220 Beta and Gamma (Deco Dosel Wholebody Ra<Lses: JUL AUG _SEE QCE NQE .DEC JAN EER MAB AER MAY JHN-ME 50 46 43 48 49 50 49 49 59 58 55 50 AME O 5 2 5 4 4 3 3 0 1 5 7 AE O 16 15 32 15 20 27 17 0 10 24 17 IIE O 30 20 20 30 .30 40 30 0 10 40 : 20 l TLD Finger Rings *: 1 JUL AUG SEE QCE HQE DEC JAN EER. MAR AEE MAY JUN- ) ME 66 70 60 62 58 67 62 67 .72 68 68 61 AME 54 50 50 54 56 52 -55 49 51 52 59 68 AE 133 158 '171 204 186 146 159 155 188 184- 189 199 IIE 710 630 900 1500 500 960 830 740 900 910- 740 1620 1 X-1 i! ! -l ! l , .1>
e
- hicludes Monthly and bi-weekly ring badges " '
SUMMARY
OF RADIATION EXPOSUIES'IO. FACIIIIY STAFF, EXPERIMENIERS AND VISTIORS 1 July 1989 through 30 June 1990 (Cont!nued) SPARE ISSUE FILM-BADGES Beta. Gamma. Neutron (Deen Dose) Wholebody Badges: JUL AUG S.EE M~ M DEC JAN EE.B MAB AEE MAX JU.R ME 51 '48 54 49 56 ' 54 34 56 -126 66 56 71 AME 5 55 1 6 30 1 21 10 14 4 1 10 AE 30 46 260 33 49 30 19 25 -27 43 20 31 IIE 70 130 260 130 110 30 50 90 90 80 -20 50 Deta and Gamma IDeen Dose) Wholebody B= lees: JUL AQQ SEE QCE NOV QEC JAN EED MAB AEB MAI ' dHN
~
ME 46 44 46 58 69 57 54 77 104 70 '56 52 AME 1 2 0 0 O. I 1 6 8 0 0 3 AE 10 10 0 0 0 10 10 33 35 0 0 20 : IIE 10 10 0 0 0 10 10 60 70 0 0 20 TLD Finger Rings: JUL AUG SEE QCI HW DEC DAN EER MAB AEB MAY JUN ME 32 97 36 94 117 87 93 52 ^ 108 91 93- 91 AME 8 40 3 27 48 28 39 31 42 22 19 20-AE 100 148 210 101 71 153 134 107 .111 89 98 104 IIE 280 760 320 890 230 580 1750 870 -870 320 290 330 SELF READING DOSIMETERS dUL AUG SEE DCI NQY DEC JAN EER MAB AEB MAI dHN ME 3 1 17 10 1 6 0 2 8 O- 6 8 AME 79 82 62 70 79 73 79 75 71 78 70 68 AE 47 55 58 .64 ~88 '57 -52 57 57 49 60 64 IIE 260 220 250 230 450 253 292 280 355 225- _235 212 '
'X-2
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