ML20052G123
| ML20052G123 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 05/10/1982 |
| From: | Sinderman R CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | |
| Shared Package | |
| ML20052G119 | List: |
| References | |
| ISSUANCES-OLA, NUDOCS 8205140328 | |
| Download: ML20052G123 (29) | |
Text
e 4
Uli!TED STATES OF AMERICA i
NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
Docket No. 50-155-OLA CONSUMERS POWER COMPANY
)
(Spent Fuel Pool
)
Modification)
(Big Rock Point Nuclear Power
)
Plant)
)
TESTIMONY OF ROGER W.
SINDEFGmN RESPONDING TO THE BOARD'S MEMORANDUM AND ORDER (CONCERNING MOTIONS FOR
SUMMARY
DISPOSITION)
My name is Roger W. Sinderman.
I am employed by Consumers Power Company as Director of Radiological Services.
My business address is 1945 West Parnall Road, Jackson, Michigan.
I joined Consumers Power Company on May 9,'1966 and-I have held various positions of increasing responsibilities since that date.
My educational background and work exper-ience are detailed in a Statement of Professional Qualifi-cations previously submitted in this proceeding.
The purpose of this testimony is to address questions raised by the Board in its Memorandum and Order i
(Concerning Motions for Summary Disposition) dated February 19, 1982.
Specifically, with respect to Christa-Maria Contention 2 and O'Neill' Contention IIA, the following concern is raised by the Board on page 32 of the Order:
" (4)
What is the reason the applicant stated that it used ' mass 8205140328 820510 PDR ADOCK 05000155 T
4-l-
absorption coefficients' in radiation estimates when it apparently used linear absorption coefficients."
The radiation estimates referred to by the Board appeared in my previous affidavit as calculations of radiation dose at the site boundary resulting from spent fuel being stored near the tapered south wall of the fuel pool.
The formula as given on page 3 of my previous testimony was taken from page 763 of the 1961 edition of Hine and Brownell (attached) and modified to take into account the shielding afforded by both the steel shell of the containment building and site boundary.
The formula is correct as given with the use of linear not mass absorption coefficients.
Linear absorption coefficients were used in making the calculations 4
and, as such, the results obtained are correct.
The term
" mass absorption coefficients" appearing in the above-mentioned testimony is incorrect.
A second area of concern raised by the Board on the j
same contention involves the ALARA program.
Specifically, on page 32 of the Memorandum and Order, the Board states, it would seem appropriate for the Director of the Radiological Services to provide his opinion on the adequacy i
of the ALARA program for the protection of workers."
Perhaps the best way to draw a reasonable focus upon the level of radiation exposure to workers (including tempo-rary workers) is to first compare the estimated radiation dose for the spent fuel pool modification with the total radiation
i, 4 ;
dose delivered to all workers during a typical year of opera-tion of the Big Rock Point Plant.
In Mr._Axtell's testimony on this subject, he states that the_ estimated dose to workers
]
for this project is 18.2 person rem.: On the'other hand, the dose delivered to workers at Big Rock Point over the last six years has averaged 290 person rem per year.
The fuel pool project, therefore, is estimated to deliver only about'6% of a yearly average dose to. workers at Big Rock Plant.
i Notwithstanding the above, it is recognized that during its evaluation report on the Big Rock Point Plant the
{
Institute for Nuclear Power Operations (INPO)'noted defi-1 ciencies in the ALARA program at Big Rock Point.
Just prior 1
4 to the INPO evaluation, Consumers Power Company began preparing a formal Corporate Radiation Safety Plan which 4
documents, among other things, the ALARA practices which are to be followed, in a consistent manner, by all of its nuclear facilities.
In addition, the position of ALARA coordinator, as required by this Plan, has been created at each of-the i
three nuclear power plants of the Company and at the General Office.
Procedures which implement the entire Radiation-Safety Plan including ALARA practices are' presently being.
I written and will be completely in effect by the end of 1982.
j This formalization of the ALARA program directly satisfies the INPO concerns.
It should be noted, however, that even in the absence of formal documentation of ALARA practices at Big Rock i
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i
. ~.
4 Point, plant practices-have traditionally taken into account-radiation exposure to workers.
While the above-cited evaluation by INPO pointed to a lack of formal documented ALARA program, examination of plant procedures including radiation protection procedures, operating procedures, and emergency procedures as well as plant practices are sufficiently representative to show that ALARA considerations have consistently been taken into account for both routine and special activities at the plant.
The placement, in individual procedures of ALARA steps and the constant ALARA prtctice review, has been the mechanism of ALARA program implementation at Big Rock Point since its early days of operation.
This traditional program has reached this point of maturity through dedicated efforts by all members of the plant staff.
Examples of procedures and practices employed by the plant to keep radiation exposure as low as reasonable achiev-able are shown in Attachments 1 through 3 of this testimony. is a plant radiation protection procedure used during emergencies to sample very highly contaminated liquid.
Section 4 and Section 6 of this procedure contain many steps to reduce radiation exposure.
For example, in Section 4, care is taken to insure that proper protective clothing is worn and sufficient water is available to dilute the sample and reduce f
its radiation field.
In Section 6, the alternatives of reducing sample volume and the use of long-handled tools are I
further examples of ALARA considerations. is an l
{
i l
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e
. operating procedure which contains several exposure
" minimization items in Sections 2 and 3.
Finally, Attachment 3 contains a recent evaluation of radiation exposure in the Steam Drum area during the 1932 refueling outage.
One result of this evaluation was a request by the plant ALARA coordinator, Mr. M. G. Dickson, to remove some no longer used plant equipment at a savings of about 5-person rem per year.
As noted, this modification was approved by Mr. Hoffman,-the assistant plant superintendent, a
on April 13, 1982.
The results of these long-standing ALARA practices can be demonstrated by comparing the total radiation dose delivered to all workers at Big Rock Point over the-last several years to that delivered to all workers at other boiling water reactor facilities in the United States over the same period of time.
The attached table (Attachment 4) entitled " Radiation Dose at Boiling Water Reactors Listed in Ascending Order of Person-Rem's per Reactor," shows such a
)
comparison for the years 1976 through 1980 for all boiling water power reactors licensed in the United States.
This information has been obtained from NUREG-0713, Volume 2 entitled Occupational Radiation Exposure at Commercial Nuclear Power Reactors 1980.
As can be seen, radiation exposures at Big Rock Point have remained consistently below the average radiation dose delivered to workers at other boiling water reacters.
This is particularly significant in light of the I
--,-a s
. fact that the larger, more contemporary reactors exhibit designs incorporating advanced ALARA features, while such features at a plant, the vintage of Big Rock' Point, are less advanced.
For 1981, the radiation dose deliverd to workers at Big Rock Point was 134 person-rem.
This dose is lower.than that recorded for the years 1976-1980 in Attachment 4.
A second method of comparison with other power reactors can be achieved by noting that during the last 14 years, exposures to radiation at Big Rock Point have consis-tently been within regulatory limits.
This comparison is valid because when ALARA practices are not followed over-exposures tend to increase.
The average exposure to workers has remained at approximately 0.6 rem / year.
This compares to an average level of about 0.75 rem / year per individual at other boiling water reactors.
Over the twenty-year operating period of the plant, only three individuals, all Consumers Power Company employees, have been overexposed to radiation.
One occurred in 1967, resulting in a dose to the skin of about 14 rem during one calendar quarter.
The limit is 7.5 rem.
The other two occurrences were in 1968, with each individual receiving a dose to the whole body between 3.1 and 3.4 rem during one calendar quarter.
The limit is 3 rem.
A comparison showing l
the overexposure records of other power reactors is shown in l'.
The data is also frcm NUREG-0713, 1980.
l t
l
s e' I While this record speaks well of the success of a consistently good radiation safety program at Big Rock Point,.
the Board's concern in this area also relates more specifi-cally to the ALARA practices that will be put into place during the modification of the spent fuel pool to accommodate expanded storage.
Mr. Axtell has, in both his testimony on this subject and.in his previous affidavit filed in connection with Christa-Maria Contention 2 and O'Neill Contentior. IIA, detailed the steps that will be taken to assure that radiation exposure to workers during this modification will. remain as alow as is reasonably achievable.
I have reviewed both the-testimony and affidavit of Mr. Axtell and am in agreement with the steps being taken to reduce exposure to workers and believe that the resulting exposures will be ALARA.
I also believe that the radiation exposures at Big-Rock Point have been consistently ALARA due to the ongoing plant practices which have evolved over a long operating history.
Finally, I believe the radiation exposure estimated for the spent fuel pool modification work is small compared to that normally i
encountered in any year at Big Rock Point.
. =.
e_
VOLUME 9A~
IMPLEMENTING PROCEDUPIS-5E ATTACHMENT 1 3IG ROCK POINT NUCLEAR PLANT SITE EMERGENCY PLAN PROCEDURE TO DETERMINE EXTENT OF CORE DAMAGE (FOR LESS THAN 10* CORE MILTDOWN)
Procedure SE 1.0 PURPOSE 1.1 To provide a procedure for an emergency monitoring technician to sample and analyze reactor water under accident conditions from the core spray heat exchanger.
Based on predicted radiation levels for various postulated accidents, personnel sampling of core spray heat exchanger water can be performed for accidents resulting in up to 10% core meltdown.
EPIP SD can be used for deter =ining 0% to 100%
i
~
core meltdown.-
- 2. 0' METHOD 2.1 A 3/4" drain line on the core spray heat exchanger will be used for obtaining the sample of primary coolant and Ge(L1) pr'ocedures will be used for the isotopic analysis.
3.0 TECHNICAL SPECIFICATIONS AND OTHER REQUIREMENTS
/
3.1 NRC commitment, memo from W L Roberts /F A Turski to distribution
( ',
list, dated November 15, 1979, entitled "NURIG-0578, Nuclear Plant Projects Listing."
4.0 SPECIAL EQUIPMENT 4.1 Volume II Procedures RP-7, " Entry Control for High-Radiation Areas" and RP-10, "Use of High-Radiation and Airborne Area Work Sheet."
4.2 Personnel and extremity dosimetry including, as necessary.TLDs, high-and low-range pocket dosimeters.
4.3 Respiratory protection cquipment, anti-contamination clothing and plastic wet suits.
l Radiological survey instruments and air sampler.
4.4 4.5 Lead shielding and extension tools for handling, obtaining and storing sample containers.
l 4.6 Print M-123, " Post-Incident Cooling System."
l 4.7 Sample and flush containers.
4.8 Demineralized water for dilution.
11/23/81 5E-1 Rev 15 l
l pr1181-2214a103
~
VOLUME 9A IMPLEMENTING PROCEDURES-5E BIG ROCK POINT NUCIIAR PLANT SITE EMERGENCY PLAN 4.9 Key for valve VPI-111, available in the Shif t Supervisor's office.
I l
5.0 PRECAUTIONS AND LIMITATIONS 15. 1 This procedure is applicable to primary coolant samp1tng during a'ccident and post accident conditions.
5.2 Procedures RP-7 and RP-10 from Volume 11 will be thoroughly reviewed by the sampling team and the Plant Health Physicist or his designated alternate.
The review shall be logged in the Radiation Protection logbook.
5.3 No person shall receive greater than 3 Rem whole body, 18-3/4 Rem to the extremities or 7-1/4 Rem skin (beta) exposure during the sampling process or subsequent analysis.
This limit includes transit to and from the sampling location.
5.4 Two persons.will be assigned to the sampling team, one to sample and one to perform radiological surveys.
'6.0 PROCEDURE 6.1 Conduct a radiological survey including air activity where necessary to and from the sampling location including the counting laboratory area. Note the readings on the area monitors.
6.2 Record the survey results in the Radiation Protection logbook and document the review of the survey results by the Plant Health Physicist (or alternate) and the sampling team.
6.3 Using the results of the survey, determine the optimum sampling methods based on current conditions.
The following should be considered in an effort to minimize exposure.
6.3.1 Determine protective equipment to be utilized, respiratory protection, anti-c clothing, wet suit?
6.3.2 Determ16e the best route to and from the core spray heat exchanger room?
6.3.3 Determine the appropriate sample volume? Under some conditions 50 ml will be necessary.
Record in Step 6.4.3.
6.3.4
, Determine the appropriate counting volume? Under high l
activity conditions a few al will be adequate.
6.3.5 Determine the optimum sample collection technique.
If the dose rate at the 3/4" drain line exceeds 4 R/hr, long handled tools should be considered.
11/23/81 SE-2 Rev 15 pr1181-2214a103
VOLUME 9A MPLEMENTING PROCEDURES-SE 3IG ROCK POINT NUCII.AR PLANT SITE E."RGENCY PLAN e
6.3.6 Locate an area to perform the volume reduction / dilution.
Under certain conditions the condensate demin area may be preferable to the core spray heat exchanger room.
6.3.7 The method to dilute or reduce the sample volume should be determined. A pipette may be appropriate if a small sample volume is desired.
6.3.8 The excess sample should be stored in a secure location.
It may be desirable to store the sample in the core spray heat exchanger room or the condensate demin area.
Consider any shielding requirements.
6.3.9 Determine how the sample should be transported to.the laboratory area.
Long handled tools or the laboratory shield cask should be considered.
NOTE:
If the background in the counting lab is greater than 10 mR/hr (making isotopic analysis meaningless), store and shield the sample for analysis off-site.'
6.3.10 Anticipate storage and dilution methods to be used in the laboratory.
[
6.4 Using the results of the survey, determine an estimate of the following:
6.4.1 Estimated whole body' individual dose received during transit from access control to the core spray heat exchanger room.
Roentgen 6.4.2
' Estimated extremity and whole body dose received while drawing sample.
Extremities-Roentgen Whole Body-Roentgen
(
NOTE:
Additional dose due to the release of noble gases from the opened 3/4" drain line may be encountered.
6.4.3 Estimated extremity and whole body dose received while transporting sample to the laboratory area.
j Extremities-Roentgen Whole 6.4.4 Estimated whole body dose received while the sample is in the laboratory area.
Roentgen
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- 11/23/S1 SE-3 Rev 15 l
pr1181-2214a103 a
- - - = =
=
VOLUME 9A IMPLEMENT NG PROCEDURES-5E BIG ROCK POINT NUCLEAR PLANT SITE EMERGENCY PLAN i
6.4.5 Based on the above evaluations, the expected doses and stay times are as follows:
Exposure Stay Times Access Control to Sample Location Draw Sample and Dilute / Reduce volume Whole Body Extremitites Return to Laboratory Dose Rate in Laboratory Dilute Sample to 4 mR/h for Spectrum Analysis Roentgen Other items reviewed and comments:
(Consider involving different teams of personnel to perform the different phases of the sample / analysis procedure.)
Prepared by f
I Reviewed by Plant Health Physicist or Alternate Reviewed by Sampling Team Reviewed by Laboratory Analysis Team 6.5 To sample the core spray heat exchanger, proceed as follows:
6.5.1 Remove the 3/4" pipe plug from VPI-111.
A tee handle is attached to ease removal.
11/23/81 SE-4 Rev 15 pr1181-2214a103
VOLLME 9A IMPII.MENTING PROCEDURES-5E BIG ROCK PODiT NUCLEAR PLANT SITE E.W RGENCY PLAN 6.5.2 Place the flush container under the drain line,. unlock chain, open the root valve and flush the sample line.
Close the root valve.
6.5.3 Place the sample container under the drain connection, open the root valve and draw off approximately al of sample.
Close the root valve, e
6.5.4 Survey the sample container.
6.5.5 Perform the appropriate dilution / volume reduction.-
Aliquot Size Aliquot Diluted Volume Reduced Initial to (mis) to (mis) 6.5.6 Transport the sample to the counting laboratory.
6.6 Document all radiological and isotopic analysis information in the Radiation Protection logbook.
7.0 LABORATORY ANALYSIS 7.1 Dilute the sample to $4 mR/h and spectrum analyze.
7.2 Determine the extent of core damage from the activity measured in Step 7.1 by using the appropriate graph on Pages SE8, SE9 or SElo.
7.3
% Fuel Melt
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l Determined by Reviewed by i
l Plant Health Physicist Site Emer~gency' Director 11/23/81 SE-5 Rev 15 p.1181-2214a103 i
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ATTACHMENT 2 Vol'?J. 3:
OPERATING PRCCEDURES - S*fSTEM SOP 44 - SPENT IUEL POOL CPERATIONS AND NEW R.1EL HANDLING 1.0 PURPOSE The Purpose of this procecure :.s to establish guidelines to be followed while working in and using equipment associated with the -
spent fuel pool.
(SFP and guidelines for the handling of new fuel.)
2.0 TECHNICAL'SPECITICATIONS AND PLANT REQUIREMENTS 2.1 TECHNICAL SPECIFICATIONS l
1.
Section 4.2.11,
2.
Section 3.6 3.
Section 7.4f,g,h 4.
Sect' ion 6.4.2b 2.2 PLANT REQUIREMENTS.
1.
klen the reactor head is off, all personnel shall be clear of the open reactor area during CRD movement.
2.
Only one fuel bundle shall be in the process of being moved at any one time in the centainment sphere except for those bundles contained in a' shipping container designed to carry more than one bundle.
3.
The fuel pool status board shall be kept current and completed moves shall be logged in the Control Room logbook.
4.
Con'tainment integrity shall be in effect when moving fuel.
, S.
A portable radiation monitoring instrument"shall be aviilable for use on the refueling deck.
i 3.0 PRECAUTIONS AND LIMITATIONS For all work that requires the use of the su bridge winch, 1.
the SFP bridge checksheet must be completed prior to use.
2.
kille using the SFP bridge winch, periodically inspect tne cable drum to insure that the cable is wrapping properly.
Whenever possible, use the mechanical stop/ cable weight and an extension cable on the vinch cable to help the cable wrap tightly.
3.
When the SFP bridge winch is to be used to move irradiated components (fuel, blades, channels', source pins) the 12/18/81 SOP 44 - 1 Rev.23 br0182-222Sa123 l
VOLUME 3:
OPERATING FROCEDURES - SYSTDi SOP 44 - SPENT FUEL POOL OPERATIONS AND NEW FUEL HANDLING mechar.ical stop/ cable weight and the 12 ft. extension cable must be installed on the winch cable along with the proper grapple for the componenet being moved. Deviation from this requirement will only be allowed if the following conditions are met:
a.
The Shift Supervisor gives permission to remove the weight and cable.
b.
The Operator shall closely monitor the ra'diation field
.as the component is lifted.
c.
A second Operator is stationed at the SFP winch breaker
,and instructed to open the breaker should the winch stick in the "up" direction.
4.
Always use a lockaht whenever the mechanical stop/ cable weight, extension cables, grapples or other hooks are attached to the SFP winch cable and periodically check all connections for tightness.
5.
Only underwater lighting equipped to contain all bulb breakage shall be used in the SFP. Never turn an underwater light fixture on unless the fixture is submerged in the pool.
~
6.
Tools should be inspected frequently, when used, to ensure that they are in operable condition. Damaged tools shall not be used and should be labeled as being damaged. A MO should be submitted promptly so repairs-can be made.
7.
Tools shall normally be used only for their intended purpose. A list posted in the reactor area and attached to this procedure indicates the approved tools to be used in I
the SFP and their intended use Any operation not listed for a. given tool (or the use of a tool not on the list) must have the approval of the Shift Supervisor on duty.
8.
Before attempting to place a fuel bundle in the fuel elevator, a functional test of the elevator should be performed.
Continous use of the elevator shall be considered a functional test'.
The elevator must be all the way down against the lower stop before a fuel bundle is placed in or removed from the elevator.
The fuel bundle must rest squarely in the bottom turntable of the elevator so it will turn freely when the locking pin is removed.
,9.
When storing miscellaneous irradiated material in the SFP, use stainless steel cables to hang the material from the SFP rail.
Place small pieces in a stainless steel bucket before l
12/18/81 S0P 44 - 2 Rev 23 l
br0182-2228a123
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VOLUME 3:
OPERATING PROCEDURES - STSTEM SOP 44 - SPENT FUEL POOL OPERATIONS AND NEW FUEL HANDLING hanging and identify all cables with a tag stating what the material is and the dose rate given off (if possible).
1 Efforts should be made to control loose parts in the SFP by containing all small parts.
10.
CAUTION should be taken not to contact the exposed 480V buses on the reactor crane when using long actuator poles in i
the SFP.
11.
If either or both of the area monitors located on the reactor deck should alarm, all personnel are to evacuate the sphere immediately by way of the personnel lock using the south stairway to leave the reactor deck. Refer to SOP-37, Step 2.1 to adjust the area monitoring alarm settings when moving components producing gamma fields greater than or l
equal to 20mR/hr, also refer to EMP 3.4, FUEL HANDLING i
ACCIDENTS.
i 12.
In order to provide safe working conditions for pe. son working in the SFP area, the following guidelihes for good l
radiological and personnel safety practices should be followed:
a.
Before entering the SFP and reactor deck areas, consult
(
the status board for the proper protective clothing and respiratory equipment to use.
l b.
Before any work is performed, make a radiological survey of the area to identify high fields.
c.
Never remove a tool, light or piece of equipment from the SFP without taking a radiation survey of the tool as it is being removed from the water.
d.
Equipment removed from the SFP should b'e decontaminated using demin water and a brush before it dries to reduce the spread of contamination and possible airborne problems.
e.
Anytime work is to be performed in the SFP, a two-man
" buddy" system will be used, each one providing a safety observer for the other.
f.
Except for emergency exit situations, never leave the reactor deck or SFP area with a fuel bundle hanging off of the SFP winch cable.
g.
No operator will be permitted to work in the SFP unless he has been qualified to work in the SFP or is working under the direct supervision of a qualified operator.
12/18/81 SOP 44 - 3 Rev.23 i
br0182-2228a123
~. - - -
VOLUME 3: OPERATING PROCEDURES - SYSTEM SOP 46 - SPENT FUEL POOL OPERATIONS AND NEW FUEL HANDLING An operator will be considered qualified to. work in the SFP if the section of the Auxiliary Operator preliminary check lirt (Volume 18, Chapter 4, Appendix C) dealing with component. handling in the SFP has been signed off.
13.
When handling new fuel in the new fuel storage area, the following guidelines are provided:
a.
Always wear clean cloth gloves when handling new fuel bundles.
b.
Operat'e the 5-ton winch in slow speed when removing or placing a bundle in a storage bay.
- c. ' An approved nylon sling may be used to attach new fuel bundles to the 5-ton hook when in the process of new fuel inspection.'
d.
Two workmen are required to remove or place a bundle in a storage bay or the inspection rack, one 'to operate the crane and one to. guide the bundle.
4.0 INITIAL CONDITIONS
(
1.
Containment integrity has been established, if fuel is to be k
moved.
2.
At least one of the two area monitors on the reactor deck j
are in service prior to handling radioactive material on the reactor deck or in the SFP, unless the containment ventilation system is isolated.
3.
The SFP bridge winch check sheet has been completed before using the bridge winch.
5.0 REFERENCES
AND ATTACHMENTS
. - SFP Bridge Check Sheet i - Approved Tool List l
SOP-2 l
50P-37 l
EMP 3.4 Technical Specifications - 6.0
' PROCEDURE 12/18/81 S0P 44 - 4 Rev 23 b r0182-2228a123 ll.
VOLUME 3:
OPERATING PROCEDURES - SYSTEM S0P 44 - SPENT FUEL POOL OPERATIONS AND NEW FUEL HANDLING 6.1 GENERAL INSTRUCTIONS FOR SFP BRIDGE OPERATION 1.
Close or check closed breaker 2P-13.
j 2.
On the SFP bridge, close both local breakers.
l l
3.
Make an inspe.ction to be sure that the bridge, winch or any tools' or lights vill not ccatact anything in the SFP or on
~
the deck if the bridge is moved.
4.
When removing or placing a fuel bundle, blade or channel in a storage rack, operate the winch in slow speed only.
l l
5.
The following table lists the controls en the SFP bridge and i
their functions:
C'ONTROL FUNCTION-Bridge forward Bridge moves east Bridge reverse Bridge moves west Trolley forward Trolley moves south Trolley reverse Trolley moves north Hoist up, half in Up slow speed Hoist up, full in Up fast speed Hoist down, half in Down slow speed
/
Hoist down, full in Down fast speed i
Bypass Allows the hoist cable to raise higher than the upper (electrical) limit Emergency stop toggle - Up Stop
- Down Run Foot brake bar Stepping on the bar stops any bridge motion 6.2 GENERAL DIRECTIONS FOR OPERATING THE GRAPPLES.
1.
Select the grapple to be used, determined b'y the component to'be moved by consulting the following list:
l l
FUEL BUNDLES - Use the short shrouded grapple.
CHANNELS
- Use the long channel grapple.
BLADES
- Use the short unshrouded grapple.
2.
Mount the chosen grapple on the winch cable that has already I
been prepared by adding the mechanical stop and extension cable.
See Step 3.3.
3.
Assemble an actuator pole and extensions long enough to reach the component to be grappled.
l 12/18/81 SOP 44 - 5 Rev 23 l
br0182-2228a123 l
l
VOLUME 3:
OPERATING PROCEDURES - SYSTEM SOP 44 - SPENT FUEL POOL OPERATIONS AND NEW FUEL HANDLING 4.
Lower.the grapple into the pool down to the top of the component to be moved and engage the grapple with the act'tator pole.
5.
Position the grapple to the spring. loaded jaws are over the lifting bail of the component.
When using the blade grapple, care should be taken to center the grapple on the length of the bail.
6.
Lower the grapple onto the bail, slacken the cable and using the. actuator pole push on the grapple to engage the bail.
Turn the grapple clockwise approximately 1/2 of a turn and hold it against the stop with the actuator pole.
7.
Using the winch, slowly tighten up on the winch cable.
If the grapple is properly engaged the grapple will turn counterclockwise approximately 1/4 of a turn as the cable is tightened.
The component may now be moved to the desired location.
8.
After placing the component in the storage rack, slacken the winch cable and engage the grapple with the actuator pole.
9.
Turn the grapple counterclockwise approximately 1/2 of a
{
turn until spring t,ension can be felt in the actuator pole.
10.
Using the winch cable, lift the grapple off of the bail, disengaging the grapple.
11.
Report the completed move to the Control Room and update the status board on the reactor deck.
Both Operators shall verify the component location and the status board.
I l
6.3 COBALT WORK l
l The procedures for cobalt rod swapping and cask loading are
' issued under a. separate title.
6.4 MISCELLANEOUS OPERATIONS When performing miscellaneous operations in the SFP, care must be taken to follow the precautions and limitations section of this procedure.
7.0 EQUIPMENT TAGGING 7.1 SFP BRIDGE AND WINCH 12/18/81 SOP 44 - 6 Rev 23 br0182-2228a123
VOLL7fE 3:
OPERATING PROCEDURES - SYSTEM SOP 44 - SPENT FUEL POOL OPERATIONS AND NEW MTEL HANDLING 7.1.1 To Remove From Service:
1.
Check that the equipment is not in use and the bridge and winch are the desired location to be worked on.
2.
Open 480V breaker 2P-13 and tag with a Red Workman's Protective Tag.
7.1.2 To Restore to Service 1.
Remove the Red Workman's Protective Tag from breaker 2P-13 and close it.
8.0 CHECKLISTS None.
12/18/S1 SOP 44 - 7 Rev,23 br0182-2228a123
. _ _ ~
FOLUME 3:
OPERATING PROCEDURES - STSTEM SOP 44 - SPENT FUEL POOL OPERATIONS AND NEW FUEL HANDLING' 4
ATTACHENT 1 FUEL POOL HOIST & BRIDGE CHECK SHEET Inspection Prior To Use 1.
Hoist cable level wound en drwn.
2.
Mechanical block installed if handling fuel, channels or materials whir.h could cause high erposure to personnel.
(
FUNCTIONAL CHECK:
a.
Raise an'd lower
. emergency stop SW.
b.
North - South travel (hoist).
c.
East - West travel.-
d.
Bridge mechanical, foot brake.
DATE JINITIAL COMMENTS I
I t
i I
t l
l i
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12/18/81 S0P 44 - 8 Rev 23 br0182-2228a123 l
i VOLUME 3:
OPERATING PROCEDURES - SYSTEM SOP 44 - SPENT FUEL. POOL OPERATIONS AND NEW FUEL HANDLING e
ATTACHMLNT 2 APPROVED TOOL LIST TOOL USE WINCH MOUNTED TOOLS i
Short Shrouded Grapple Lifting fuel bundles Short Unshrouded Grapple Lifting control rod blades Long Channel Grapple Lifting channels
{
Pin Grapple (Double J Hook)
Lifting source pics or cans i
Incore Grapple Lifting incore detector strings Single Safety Hook Miscellaneous lifting not listed abose Treat II Can Handling Tool Assist placing Treat II Can,into cask HAND TOOLS Actuator Pole Operating grapples 1, 2 and 3 abovs Bodendyke Pole Operating pole for various tools designed
\\
to mate with it l
Switch Pole Assist in operating other tool
{
minor manipulations Extension Pole Extend the length of the above poles Bodendyke Vise Grip Pliers Miscellaneous gripping l
Bodendyke Bronze Grippers Miscellaneous gripping Bodendyke Dum Dum Holder To use Dum Dum to pickup small parts or identify fuel bundles Bodendyke Retrieval Fingers Pickup small parts Bodendyke Wedges Prying I
Bodendyke Source Can Inspection Gage Measure source cans l
Bodendyke Hack Saw Sawing - Special permission from l
Shift Supervisor Bodendyke C'-Clamps Miscellaneous clamping l
l l
12/18/81 SOP 44 - 9 Rev 23 br0182-2228a123
UOL12E 3:
OPERATfNG PROCEDURES - SYSTEM SOP 44 - SPENT FUEL POOL OPERATIONS AND NEW FUEL HANDLING i
TOOL USE Hammer Pole Tapping, Hammering Plexiglass Win'dow Viewing aid Pneusatic Pliers Gripping COBALT AND SOURCE PIN TOOLS i
l NOTEi All tools below are a single 15 foot pole with the tool mounted l
on the end.
Forked End Tool "F" fuel source and cobalt ro,d removal 5/8" Hex Socket "F" fuel cobalt rod unlocker Large Finger Tool "F" fuel cobalt rod lifter NPI Removal Tool "F" fuel cobalt rod removal' 5/8" Square Socket "F" fuel source pin removal l
1/2" Hex Socket "G" fuel cobalt rod unlocker 1/2" Square Socket "G" fuel source pin removal' Small Finger Tool "G" fuel cobalt rod lifter Side Grabber All rods lifter Snare All rods lifter l
1 I
12/18/81 SOP 44 - 10 Rev 23 br0182-2228a123
ATTACHMENT 3 To DPEof.^..an, ?2h-61h3 ICDickson[b' Foo ConSum8tS o,7c April 13, 1982 SQWg[
susacet I!G EOCK POINT PLANT-S!T.AM DRUM AP.EA IIPOSURE-ALARA
'.NTCRNAL CorngsrcNogge MGL 82-lh cc CIAxtell DDHerboldshei=cr RESchrader JL?cntaine, P2h-20h l
ACSevener DC 2h*01*02 JJ?opa A review of exposure in the Steam Drun area during the 1982 outage indicates thfee =ethods to =ini=ize exposuret
- 1) Ti=e-A vell planned activity is the =est effective =ethod to reduce the ti=e required to perfor= verk. Generally vo-k vas perfer=ed efficiently, however, aggressive job planning has the greatest po-tential to pay exposure dividends.
2 ).
Shielding-The best and only reasonable =ethod to shield the stea=
dru= area is internally vith vater.
Efferts,_ should be =ade to =ex-i=1:e the ti=e that the dr..~ and asscciated piping is filled with vater.
In addition, Job scheduling should be reviewed to deter.ine the exposure benefits of perfor=ing various tasks during shielded ~
periods.
- 3) Wait Area-The stea= drun access area (inside) dese rates are presend 17 up to 75 =R/hr and increasing each outase. The occupancy time l
(100-150 ranheurs per year) is significant since this area is used for toe rubbers and a staging ares to *
'-d e exposure of verkers and technicians.
It is anticipated that the dese rate could be reduced to less than 25 =R/hr by re=oving the obsolete level gage lines (the =ajor scree) and possible installation of lead on the floor grating.
The expose e sayings, although not specifically identifiable are expected to be in l
excess of 5 =anre= per year.
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FORM RFM Consumars Povar Co=pany Nuclear Plant I,
REQUEST POR MODIFICATICN
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Plant:
8/ 6$
A C C <-
d
/o/NT G.O. Project Engr:
Proposed ModiELcation:' /CEMo vE rHC r su o e nsa c tri / % s r g,4 pr DeaM t r a.C t.
c ax s. c t wes
, fab ho-^/oo'?
4 eB swo w s.)
o A./
.D4Awi A/ 6 i -r -.4 7. 5 _
s rRusN me e s u es px s o A.
ro woAx si./b ' p awni z c' s 7 CAM DeaM ry s8 c ".h s.
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Reason for ModLfication: rHe c iu es
--a rar.ws o A. obrAA7 ak: Aec DcAb s.e'es Aub wo.rc / Arcs HAst /4ce6Asch 7o C'a o a f/A, f'c o u rae r-}.
/?fH e,o, of TNCsr c o u CS a ast L A* Eb u t e f.epolare E M Y' 5" RAud d'M,JCA yrAk j u_h usdv Akonis o /~
- s, ca a ws,
e pit. c
,cuaas i utrxe caro s u of
,4GS DJ4S M -$~ 7f P'.M -/ ?- J lG.O USE OMI.,Y 3equested By:
- [ #
Sect.'on Head Acproval:
Date:
7//J//.2.--
Mas. P-II Approval:
I l PLANT USE ONT Y l l
Approved l l
Disapproved l l Plant Seperintendent:
Date:
l Offsite Support Requested Priority 1
2 3
YES NO i
l PLANT USE ONLYj Facility Change No:
Secpoint Change No:
i
- l Cor.nents:
Plant Responsible Individual:
[
Assigned By:
Date:
I e
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RADIATIOli DOSC AT ATTACllMENT 4 DOII ING WATER REACTOftS
~'
~
1.lSTED IN ASCENDING ORDER OF PERSON-REMS Pl:R REACTOR l
1 erson-I330 r
~
~
I 1926 Peison-1977 1rerson-1978 1 erson-1979 P
P Dems Rems Rems Rems per ocr pur per pcr Site Name Sgte N11"..N'.i me Site Site Name Site Site Name Site Site Name S i t e___
Igumswidt Day 22 Duana Arnold 105 Cooper Station 198 Cooper Station 158 Ilumboldt Day 31 1,acrosso 218 I"# N 80 110 Lacrosse 225 Lacrosse 164 tunticello 157 Itatch 162 449 Drown Ferry Vermont Yankee 258 Dig Rock Point 175 Lacrosso 186 16 234 Dig Rock Point 354 Duane Arnold 299 ilatch 1 248 Cooper 221 Itatch 134 Monticello 531-Dig Mock Point 334 Nine Mile Pt 314 Duar.e Arnold 275 Fitzpatrick 202 Hine Mile Pt 591 Millntone Pt 1 394 Itumboldt Day 335 Dig Rock Point 455 rionticello 263 m m Ferry Drowns Ferry 1&2 863 Vermont Yankee' 339 Oyster Creek 4F7 1' 263 1825 Dies Rock Point 289 Itatch 1 465 Monticello 375 Drowns Ferry Duane Arnold 671 Drunswick 2 326 1, 263 1667 Ouad Cities 1&2 1031 Brunswick 1&2 1004 Dresden 1, 263 2105
(,ooler Station 350 11atch 282 Vernont Yankee 411 Dre sde n 1, 2 & 3 1000 C UP'I Monticello 1000 Drowns Ferry Peach Bottom Peach Pottom 1, 263 1792 Peach Dattom 263 2302 2&l 840 Peach Dottom 2&3 1388 2&3 2036 Peach Bottom Verinnnt Yankee 1338 Nine fille Pt 428 263 1317 Fitzpatrick B59 Fitzpatrick 1080 Oyster Creek 1733 Drenden 1, 2&l 1680 Quad Cities 1&2 1618 Pilgrim 1015 Drunswick 2 1120 Drunswick '*, &2 3870 Ilumbod i t Day 683 Fitzpatrick 909 Quad cities 162 2158 Nine Mile Pt 1383 Fitzpatrick 2040 Quad Cities 1&2 1651 Duane Arnold 974 Vermont Yankee 1170 Oyster Creek 1614 Millstone Pt 1 2158 Oyster Creek 1078 Millstone 1 1239 Drunswick 152 260J ilumboldt Day 1905 Quad Cities 1&2 4838 Millntone 1 1194 Oyster Creek 1279 Nine Mile Pt 1497 Pilgrim 1 3142 Pilgrim 3626 Pilgrim 1 2648 Pligrim 1327 Millstone Pt 2 1793 Avgs per Aveis per Reactor ft 2 R Avgs per Avgs per ctor 1136 Reactor 547 Reactor 604
. Reactor 773 I
t hose ni ten wi th more tlian onn operating reactor, the namlw'r of Foi g
person-temn poi reactor in obtained by dividing the nismler of Imrson-rems & port ed by t he site by the number of reactors.
?-
ATTACFF.ENT 5 OVEREXPOSURES TO RADIATION AT NUCLEAR POWER REACTORS Number of Workers Maximum Overexposed to Whole Body Year External Radiation Dose (Rems)
J 1371 2
3.1 1972 16 5.1 1973 19 4.0 1974 43 6.1 1975 14 3.8 1976 20 10.1 1977 27 3.6 1978 9
27.3
~
1979 21 10.1 1980 73 4.9 During this time period, none of these overexposures occurred at the Big Rock Point Plant.
l i
)
)
. -. - -.