ML20235G831
| ML20235G831 | |
| Person / Time | |
|---|---|
| Site: | 07003013 |
| Issue date: | 09/30/1985 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20235F489 | List: |
| References | |
| PROC-850930-01, NUDOCS 8707140397 | |
| Download: ML20235G831 (102) | |
Text
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COMMolNEALTH EDISON RADIATION PROTECTION STANDARDS
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8707140397 870506 September 1985 ADDCK 07003013 PDR PDR C
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i September 1985 IABLE OF CONTENTS Pace Number _
Revision Dat Radiation Protection Standards Administrative controls i
Preface 1 Feb. '82 2
Feb. '82 Definition of Terms 5
Feb. '82 Work in Controlled Areas 9
Feb. '82 Radiological Signs and Labels 11 Sep. '85 l Radiation Work Permit 18 Feb. '82 Radiation Occurrence Reports Personnel External Exposure Controls Personnel Occupational External Dose Limits 24 Feb. '82 Personnel Exposures Under Emergency Conditions 26 Feb. '82 Removal and Return of Personnel from Radiation Work 28 Feb. '82 30 Feb. '82 Dose Rate Determination 32 Feb. '82 Personnel Dosimeters Occupational Radiation Exposure Received at i
35 Feb. '82 j of f-site Locations Personnel Internal Exposure Controls i
37 Feb. '87 l Respiratory Protection 39
.Feb. '87 l controlled Area injuries Welding, Flame Cutting, Grinding or Heating 41 Feb.
'8*
Radioactive materials 42 Feb.
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Personnel Bioassay Sampling Frequency Contamination controls 45 Feb. '8' Radiological Protective Clothing Personal Effects Permitted in Contaminated Areas 47 Feb. '8 48 Sep. '8 Unconditional Releases Transporting Radioactive Materials within the 50 Feb. '8 Restricted Area 52 Feb. '8 Storage of contaminated Materials Shipment and Receipt of Radioactive Materials 5$
Feb. '8 Off-Site Shipment of Radioactive Material 56 Feb. '8 Receipt of Radioactive Materials from Off-Site
1 September 1985 TABLE OF CONTENTS (CONTINUED)
Page Number Revision Dat Radiation Protection Standards i
specific controls Collection. Transfer and Handling of Radioactive materials in the Radiochemical Laboratory and 57 Feb. '82 Counting Rooms 59 Feb. '82 Radioactive Sources Movement of Radioactive Materials in the Storage 61 Feb. '82 and Transfer Pools 62 Feb. '82 Laundry / Dry Cleaning Facility 64 Feb. '82 !
Sandblasting Cave 65 Feb. '82 Personnel Radiation Protection Training 66 Sep. '85 References 9
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o February, 1982 PREFACE Purpose The purpose of the Radiation Protection Standards is to provide guidelines for the promulgation of radiation protection ' procedures and policies at eacn of Commonwealth Edison's (CECO. 's) nuclear facilities.
It i s, the policy of Commonucalth Edison Company to maintain occupational radiation exposures as low as is reasonably achievable ( ALARA), consistent with station construction, maintenance, and operational requirements, and with economic and social considerations being taken into account.
The Commonwecith Edison Company Radiation Protection Standards are intended to reflect CECO.'s management commitment to this policy.
General In addition to prescribing minimum acceptable radiation protection practices, these standards assert that it is the responsibility of the station Radiation-Chemistry department at each nuclear f acility to ensure that procedures and policies are consistent with all applicable regulatory requirements intended for protection of individuals against radiation.
I Notwithstanding the responsibilities of each f acility's health physics staff, it remains the responsibility of each individual to keep his radiation exposure as low as is reasonably achievable.
4 from these standards must be promptly reported to the Deviations Radiation-Chemistry Supervisor and the Shift Engineer and should be documented by a Radiation Occurrence Report.
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Questions regarding an interpretation of technical aspects of material i
contained in this manual should be directed to the Radiation-Chemistry Supervisor.
Radiation Protection Standards have been prepared as a unit and the
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The individual standards, therefore, are not independent of each other.
Page 1 of 69
February, 1982 DEFINITION OF TEP,MS, S
Purpose The purpose of this section is to define those terms referenced in the Radiation Protection Standards.
Terms Used to Designate Whole Body Doses 1.
Daily whole Body. Dose:
That whole body dose' equivalent received in the twenty-four hour period commencing with the start of the worker's shift.
1 2,
Weekly Whole Body Doses That whole body dose equivalent received during the seven day period commencing at 0001 Monday and terminating at ' 2400 J
Sunday.
3.
Quarterly Whole Body pose:
That whole body dose equivalent received during the calenet;.r quarter which. is not less - than 12 consecutive weeks nor more than
'4 consecutive weeks; the first calendar quarter of each year begins ta January with subsequent calendar quarters following such day is included in more than one calendar quarter or omitted that no 0ne from inclusion within a calendar quarter.
4.
Annual Whole Body Dose:
That whole body dose equivalent received during the year's four calendar quarters.
Terms Utsed in the Designation of Areas 1.
Restricted Areat Any area to which access is controlled by the licensee for purposes of protection of individuals f rom exposure to radiation and-radioactive materials.
The restricted area is further classified as follows:
(a)
Controlled Area Any area within the restricted area requiring radiation posting.
(b)
Uncontrolled Areas Any area within the restricted area not requiring radiation posting.
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l February, 1962 DEFINITION 07 TERMS (CONTINUED) 2.
Unrestricted Areat Any area to which access is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials.
Permissible levels of radiation in the unrestricted area 2ssure thats (a)
No individual shall receive a dose equivalent to the whole body in any period of one calendar year in excess of 0.5 rem; and (c)
No individual, 'if continuously present in the area, could receive a dose equivalent to the whole body in excess of two millirem in any one hour or a dose equivalent in excess of one hundred millirem in any seven consecutive days.
3.
Radiation Areat 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 hour a dose in excess of 5 millirem, or in any 5 consecutive days a dose in excess of 100 millirem.
4.
High Radiation Areat Any area accessible to personnel in which there exists radiation at such levels that a major portion of tne body could receive in any one hour a dose in excess of 100 millirem.
5.
Airborne Radioactivity Arear
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(a) Any room, enclosure, or operating area in which airborne radioactive materials, composed wholly or in part of licensed material, exists in concentrations in excess of the amounts specified in 10 CFR 20.
(b')
Any room, enclosure, or operating area in which airborne radioactive materials composed wholly or partly of licensed material exists,in concentrations which, averaged over the number of hours in any week during which individuals are in the area, exceed 25 percent of the amounts specified in 10 CFR 20.
6.
Radioactive Material (s) Areat Any area or room in which radioactive materials are used or stored and which contains any radioactive material in an amount exceeding the quantity specified in 10 CFR 20.
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February, 1982 1
DEFINITION OF TERMS (CONTINUED 1 I
Radiation Protection Terminology 1,
Shall (will, must) : Denotes a mandatory requirement.
2.
Should:
Denotes a recommendation which is to be applied whenever practical.
3.
May Denotes permission; neither a requirement nor a recommendation.'
Terms Used in the Designation of Personnel i
The use of upper case letters in a title refers to a specific position and person, or that person's designated alternate in his absence.
The use of lower case letters in a title refers only to that position.
Radiation-Chemistry:
Cognizant personnel in the Radiation-Chemist 1y I
Department who have the, knowledge and authority to make health physics decisions as defined in these standards.
Note:
In this document Radiation-Chemistry and the' Radiation-Chemistry department are used inte rcha ngeably. This term includes i
(a)
The Radiation-Chemistry Supervisor (b) All station Health Physicists (c) All station Radiation-Chemistry Foremen (d) All station Radiation-Chemistry Technicians Raoiation-Chemistry Supervisor This is the individual at the station that has the title Radiation-Chemistry Supervisor or his designated alternate, i.e.
the Shift Engineer if both the Radiation-Chemistry Supervisor and the Lead Health Physicist are not on site and cannot be contacted.
Lead Health Physicist (Station Health Physicist) a nd Health Physic M Management personnel with the minimum of a four year college degree in a science or engineering field with some formal training in Health Physics, l
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i Feoruary, 1982 WORK IN CONTROLLED AREAS Purpose This standard describes the radiological requirements for working in controlled areas such that:
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internal radiation exposures to personnel are maintained as 1.
External and low as is reasonably achievable (ALARA).
The spread of radioactive materials or contamination is minimized.
2.
j General 1.
All individuals working in or f requenting any portion of a controlled area shall have been trained in accordance with 10 CFR Part 19 prior to entry the controlled area.
Certain trained individuals may also be into (for required to be accompanied by a duly trained person in order to visit purposes of inspection or tours) any portion of a controlled area.
be authorized by 2.
All work performed by personnel in controlled areas must supervision.
All entries to high radiation areas shall require authorization by the Shif t Engineer or the operating unit shift supervisor.
3.
Each entrance or access point to a high radiation area must have access controls which comply with the requirements of 10 CFR 20.
The supervisor responsible for performing the work shall determine that 4.
each individual entering the area understands the radiological precautions necessary for performing the assigned duties in a safe manner and should ensure that the individual is aware of his current accumulated radiation cose and the maximum dose equivalent authorized for the job.
5.
Notwithstanding the ef forts and assistance off ered by the health physics staff, each individual performing work in a controlled area must be aware for the that it is his responsibility to maintain his radiation exposure specific job as low as is reasonably achievable.
Each individual should be aware of the dose equivalent NOTE:
received for each work assignment.
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February, 1982 WORK IN CONTROLLED AREAS (CONTINUED)
Personnel Conduct Prior to Entering a Controlled Area 1.
The job supervisor and/or the individual shall obtain the following information from the Radiation-Chemistry groups Tne dose rates and monitoring requirements for the job, a.
b.
Protective clothing requirements.
Respiratory protection requirements.
c.
d.
Timekeeping and dosimetry requirements.
e.
Special precautions due to unusual radiological conditions noted or suspected.
2.
The approval of the Radiation-Chemistry Supervisor or the Lead Health Physicist must be obtained prior to working in areas where whole body working dose equivalent rates are greater than 3 rem / hour.
3.
Each worker shall inform the Radiation-Chemistry department of any open wounds or other skin injuries or skin conditions which might require special protective measures to prevent contamination of the wound.
Personnel conduct in a Controlled Area Each individual who performs work in a controlled area shall carry out h'is 1.
assignment in a manner which will assure that his radiation exposure as well as the radiation exposure of fellow workers is as low as is reasonably achievable.
2.
When radiation work involves raising radioactive materials in the fuel pools above established limits, uncovering contaminated materials, opening contaminated process lines, or disassembling potentially contaminated equipment where dose rates or airborne radioactivity may be expected to increase sig nificantly, the Radiation-Chemistry Department shall be informed and/or consulted before the fact so that a
radiciogical evaluation can be made.
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February, 1982 WORK IN CONTROLIID AREAS (CONTINUED) 3.
Minimize exposure and the spread of-contamination while in controlled areas by observing the following rules that all surfaces and objects are contaminated unless radiation a.
Assume surveys indicate that contamination is not present.
b.
Perform all work in such a manner that the possibilit$ of spreading contamination is minimizec.
c.
Do not eat, smoke, drink, or chew, etc. in controlled areas.
d.
Avoid local " Hot Spots",
radiation beams and " casual" or other non-work related exposures such as loitering in or around controlled areas, and utilize time, distance, and shielding protective methods whenever practical.
Observe precautions noted on all radiological signs and labels.
e.
4.
If it is suspected ~ that the radiological conditions have worsened or may change to such an extent as to lead to abnormal radiological conditions, leave. the area.
immediately and notify supervision and the Radiation-Chemistry department.
Personnel Conduct When Leavino a Controlled Area 1.
A worker should leave the controlled area as quickly as possible,.
consistent with safety, for any of the following reasons:
a.
When instructed or signaled to do so by the Radiation-Chemistry department.
b.
Failure or suspected failure of personal protective equipment.
Unexpected deterioration of radiological conditions.
c.
d.
In the event that the worker's current accumulated dose equivalent status becomes uncertain for any reason or dose equivalent is equal to the exposure authorized for the job,
" Assembly" sirens sound - practice or actual.
e.
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February, 1982 WORK IN CONTROLLED AREAS (CONTINUED) f.
Completion of work assignment.
g.
In]ury h.
Unexpected area radiation monitor alarm and the area dose rate is unknown.
2.
A worker may return to work in a controlled area when his return is approved by the Radiation-Chemistry department.
3.
When the " assembly" sirens sound (other than for an announced drill) :
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a.
Normal protective clothing removal procedures shall be waived.
l Personnel shall remove at least gloves and protective footwear when
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exiting from the controlled area.
b.
Hand and shoe counts and normally required personnel surveys may be omitted.
c.
At the first opportunity, personnel contamination checks shall be performed.
If significant contamination is discovered the I
Radiation-Chemistry department shall be notified as soon as possible.
d.
Personnel exiting from controlled areas wearing protective clothing during an " assembly" should make every reasonable effort to avoid contaminating equipment, walls, floors, and other personnel.
4.
Personnel shall leave the controlled area in a manner consistent with' station procedures and any specific instructions which may have been issued by the Radiation-Chemistry department.
S.
In case.of injury; surveys, step-off pad procedures, and efforts to prevent spreading contamination should be performed only to the extent that urgent medical treatment is not delayed.
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February, 1962 RADIOLOGICAL SIONS AND LABELS 1
Purpose i
This standard establishes the proper use of radiological signs and labels as
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required by state and federal regulations, j
General 1.
All radiological signs and labels shall have a yellow background with magenta or purple lettering.
At least one conventional magenta or purple colored three-bladed raciation symbol must appear on each sign in l
accordance with 10 CFR 20.
2.
Each controlled area shall be conspicuously posted in such a manner so as to warn personnel approaching the area from any direction.
3.
Signs and labels that warn of existing radiological conditions shall be promptly removed or updated when no longer appropriate.
4.
The use of the word " Danger" shall be li[nited to high radiation areas only.
The word " Caution" is to be used on all other radiological signs and labels.
Sions 1.
All controlled areas must be conspicuously posted with the appropriate sign or signs as definec in 10 CFR 20.
The various signs shall bear the radiation symbol and the following words:
a.
" Caution - Radiation Area" b.
" Danger - High Radiation Area" c.
" Caution - Radioactive Material" d.
" Caution - Airborne Radioactivity Area" NOTE:
If applicable, combinations of the above wording may be used on the j
same sign.
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Fe bruary, 1982 RADIOLOGICAL SIGNS AND LABELS (CONTINUED) 2.
In addition to the contents of signs prescribed in this section, additional information may be provided on or near such signs in order to aid an individual in minimizing his exposure to radiation or radioactive materials in accordance with 10 CFR 20, as for example:
Storage Area, Contaminated Floor, Authorized Entry Only, Hot Spot, etc.
All areas in which radioactive material is present in uncontained and readily 2 for alpha emitters and dispersible form in excess of 2,200 dpm/100 cm 22,000 dpm/100 cm for beta, gamma emitters shall, in addition to 2
" Caution-Radioactive Material," be conspicuously posted with the words:
Contact Radiation-Chemistry
" Potential Airborne Radioactivity Area Before Entry" Labels 1.
All labels for radioactive material packages and any other labels used to provide additional information to workers must be such that the requirements as specified in 49 CFR 173 and 10 CFR 20 are met.
In addition to the contents of labels prescribed in this section, 2.
ad6itional information may be provided on or near such labels in order to aid an individual in minimizing his exposure to radiation and radioactive material; for example:
isotope, activity, dose equivalent rate, date.
Page 10 of 69 4
September 1985 RADIATION WORK PERMIT TYPE 1 PURPOSE This standard describes those situations which require a Type 1 Radiation Work permit (RWP) and describes ~the steps necessary to obtain and complete the RWP.
CONDITIONS REOUIRING A TYPE 1 RADIATION WORK PERMIT A Type 1 RWP is required for ALL routine access or work in radiologically controlled areas where personnel are NOT expected to exceed a whole body-dose equivalent of 50 mrem / day.
DEFINITIONS f
1.
Changing conditions:
Any change in the radiological conditions that would alter the protective equipment or
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timekeeping requirements stated on the RWP.
2.
Deactivation:
The temporary suspension of an approved RWP, due to changing conditions, until the area covered by that RWP is returned to the conditions stated on the RWP.
l Any task that has a minimum number of static 3.
Routine:
whole body dose rates that are less thgn 100 arem/hr.
A 4.
Operating Supervision:
Shif t Engineer or Respective Operating Unit Shift Supervisor. For Rad. Waste work, this may be the Rad. Vaste Foreman.
GENERAL Each Type 1 RWP consists of three forms. The first is a standard RWP 1.
The form setting forth conditions and requirements for that RWP.
second is a Type 1 RWP Approval Log for Radiation-Chemistry Foreman The third is a approvals or periodic operating reviews of that RWP.
Type 1 RWP Signature Log for personnel working under the provisions of that RWP.
I If A T' pe' l RWP is valid for a maximum of 1 year from January 1.
2.
y conditions change during the course of that year the RWP shall be deactivated or terminated at'the discretion of Radiation-Chemistry.
Radiation-Chemistry Supervision shall notify the proper work group (s)
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of the change in status of any RWP.
Each Type 1 RWP shall be initially. completed by Radiation-Chemistry.
The Radiation-Chemistry Foreman shall review the ALARA Check List and 3.
RWP and sign and date the
- Rad. Chen, Approval" section prior to any other individual signing that RWP. He shall also review and approve all Type 1 RWP's on page 2 (Type 1 RWP Approval Log), once a week.
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September 198b
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4.
The ALARA Coordinator or his designated alternate shall attach a' written ALARA Action Review and sign and date the "ALARA Action I
Review" section as required by station procedures, i
5.
Operating Supervision shall read and understand all Type 1 RWP's.
He j
shall make a " Periodic Review" determination and indicate the
" Frequency" of review, or state the " Reason" if a periodic review is NOT required, and sign and date the " Operating Approval" section. If a review is required, he will sign and date on page 2 (Type 1 RWP-Approval Log) as " Review Frequency" requires.
6.
PRIOR TO ENTERING the controlled area and the FIRST time each individual works under a specific RWP; he'shal1~ read, understand and j
sign the RWP at the Radiation-Chemistry office or' designated Radiation-Chemistry office..He shall also enter:his film /TLD badge number and the date on page 3 (Type 1 RWP Signature Log).
Each individual shall comply with the requirements of the RWP in all respects. He should also review a11' current surveys of the area.
NOTE:
PRIOR to activating a Type 1 RWP, the RWP and page 3's may be brought to the various departments or work groups by an.
RCT, where the RWP will be explained and the workers allowed to sign on page 3.
"l.
Weekly whole body dose equivalents in excess of 300 mrem must be approved by the Radiation-Chemistry Supervisor or by a health physicist designated by him in writing before the Radiation-Chem!stry Foreman can authorize department approval.
8.
A Type 1 RWP shall not be used for access to, or work in High Radiation Areas, except for routine operator High Rad. rounds.
Except for routine access RWP's, each department or work. group shall 9.
work under it's own RWP, designated for that group and their " support-group (s)" only.
(See RWP instructions)'
10.
Radiation-Chemirtry shall determine when a Type 1 RWP is used.
- 11. ' Each Type 1 RWP shall be uniquely numbered by Radiation-Chemistry.
- 12. Type l RWP's shall have an initial survey prior to the start of work on that RWP and all active Type 1 RWP's shall be resurveyed at a frequency specified by station procedure (s). Surveys shall also be made upon the request of the worker.
13.
Survey data shall be kept readily available in or near the Radiation-Chemistry office (s) for worker review.
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September 198b i
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The original pages of all activated Type 1 RWP's shall be'kept in.a conspicuous location in the Radiation-Chemistry office (s).
Copies of these forms may be taken to the job site provided they are marked and
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dated as copies.
- 15. The RWP may be terminated for any of the reasons stated on the bottom of page.1 of the RWP. Whoever terminates the RWP shall sign and date and indicate the reason for termination. He shall also notify Radiation-Chemistry and ALL work groups involved with that RWP.
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September 1985 RADIATION WORK PERMIT TYPE 2 i
PURPOSE This standard describes those situations which require a Type 2 Radiation Work permit (RWP) and describes the steps necessary to obtain and complete the RWP.
CONDITIONS REOUIRING A TYPE 2 RADIATION WORK PERMIT A Type 2 RWP is required for ALL access or work in radiologically controlled areas where personnel are expected to exceed a whole body dose equivalent of 50 mrem / day.
In addition, jobs involving significant contamination and/or airborne radioactivity may require a Type 2 RVP as specified by Radiation-Chemistry Supervision.
DEFINITIONS 1.
Changing Conditions:
Any change in the radiological conditions that would alter-the protective equipment.or timekeeping requirements stated on the RWP.
2.
Deactivation:
1.
The temporary suspension of an approved RWP pending supervision approvals and worker intitials.
2.
The temporary suspension of an approved RVP due to changing conditions until the area covered by that RWP is returned to the conditions stated on the RWP.
3.
operating Supervision: Shift Engineer or Respective Operating Unit Shift Supervisor. For Rad. Vaste work, this may be the Rad. Vaste Foreman.
G_ENERAL Each Type 2 RVP consists of two forms. The first is a standard RWP 1.
The form setting forth conditions and requirements for that RWP.
second is a Type 2 RWP Approval and Signature Log for supervisory approvals and. signatures of personnel working under the provisions of that RVP.
If conditions A Type 2 RVP is valid for the length of the job.
2.
change during the course of the job, the RWP shall be deactivated or terminated at the discretion of Radiation-Chemistry.
Radiation-Chemistry Supervision shall notify the proper work group (s) of the change in status of any RVP.
September 1985 0 )
1 The requesting department or work group shall initially complete.the 3.
top portion of each Type 2 RWP. They shall then submit the RWP to Radiation-Chemistry for further action.
Radiation-Chemistry shall then perform and attach the required 4.
The. Radiation-Chemistry Foreman shall surveys and ALARA Check List.
review the ALARA Check List and RWP and sign and date-the " Rad. Chem.
Approval" section.
(see RWP instructions)
The ALARA Coordinator or his. designated alternate shall attach.a 5.
written ALARA Action Review and sign and date the "ALARA Action Review" section as required,by station procedures.
Operating Supervision shall read and understand all Type 2 RWP's.
He 6.
shall make a " Periodic Review" determination and indicate the
" Frequency" of review, or state the " Reason" if a' periodic review isIf NOT required, and sign and date the " Operating' Approval" section.
a review is required, he will initial on page 2 (Type 2 RWP Approval and Signature Log) as " Review Frequency" requires.
PRIOR to any worker signirig on a Type 2 RVP. the following 7.
information shall be completed on page 2 of the RWP.
The RCT shall enter the present date and time of NOTE:
A.
activation.
The Radiation-Chemistry Foreman shall then' read and 3.
understand the RWP and initial his approval for that date and shift.
Operating Supervision, if required, shall then read 3
c.
and understand the RWP and initial their approval for that date and shift.
The Job Foreman shall read and understand the RVP.
D.
then sign and initial for the first day and shift his crew is working. He must initial on each day of that week his crew works under that RWP.
PRI'OR'TO STARTING WORK on a Type 2 RWP, each individual shall read, 8.
understand and sign the RVP at the Radiation-Chemistry office or designated Radiation-Chemistry office on page 2 (Type 2 RVP Approval and Signature Log). He shall also enter his film /TLD badge number.
On the date he works, he will enter his approved daily dose and Each individual shall comply with the requirements of the initials.
He should also review all current surveys of RWP in all respects.
the area.
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September 1985 For daily whole body dose equivalents in excess of 100 arem, specific 9.
approval must be obtained for each job from the' Radiation-Chemistry Supervisor or from a health physicist designated in writing by him before the Radiation-Chemistry Foreman can authorize department approval.
Weekly whole body dose equivalents in excess of 300 mrem must be-10.
approved by the Radiation-Chemistry Supervisor or by a health physicist designated by him in writing before the Radiation-Chemistry Foreman can authorize department approval.
A Type 2 RVP shall be used for access to, or work in High Radiation 11.
Areas, except as defined in General 88 of Type 1 RWP's.
Each department or work group.shall work under.it's own RWP, 12.
(see designated for that group and their " support group (s)" only.
RWP instructions)
Radiation-Chemistry shall determine when a Type 2 RWP is used.
)
13.
Each Type 2 RVP shall be uhiquely numbered by Radiation-Chemistry.
14.
Type 2 RVP's shall have an inititsi survey prior to the start of work 15.
The work area dose rates shall be verified at least on that RVP.
The work area of active RVP's once a day whi'le work is in ptogress.
shall be resurveyed at least once a week while work is in progress.
Surveys shall also be made upon the request of the worker, i
Current survey sheets shall be attached to all active Type 2 RWP's.
16.
in a The original pages of all activated Type 2 RWP's shall be kept 17.
conspicuous location in the Radiation-Chemistry of fice(s). Copies of these forms may be taken to the job site provided they are marked and dated as co?ies.
The RWP may be terminated for any of the reasons stated on the bottom 18.
Whoever terminates the RWP shall sign and date of page 1 of the RVP.
and indicate the reason for termination. He shall also notify Radiat, ion-Chemistry and A1.1. work groups involved with that RWP.
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RADIATION WORK PERMIT RWPs TYPE.
S7As
'VORK DESCRIPfl0N JOB MAN HOUR EST j
JOB CATEGORY CODE
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WORK LOCAtl0N
- NA
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C NA EPN.
4 C PROCESS LINES TO BE OPENED W R.s C 10 INVOLVE WELDING. FLAME CUTTING. GRINDING OR HEATING CONT AMINATED MATE
"" 70 INVOLVE SPLASHING OR RUhNING WATER C VISUAL INSPECTION ONLY C NA SUPPORT GRCUPS DATE REQUESTOR DEPT PHONEe work AREA suRytt REsuLis f
C SURVEY AT START OF JOB WITH RCT IN l
2 SURVEY PRIOR TO JOB CONTINUAL ATTENDANCE AT JOB SITE PRE J0Et SURVEY 00$E RATE SMEARABLE f
. (mrtm mrad /hr)
(corn 100cm2)
LDCATION (mheNE (mremMr)
EST. WHOLE BODY DOSE RATE EXTREMITY DOSE FU.TE.
TEMPORARY SHIEL0 LNG INSTALLEDYES C NO '
A!RBORNE RADI0 ACTIVITY ALARA ACTION REVIEW REQUIRED YES : NO JOB MAN. REM E3T PROTECTIVE EQUIPMENT REQUIRED D051METMY RESPIRATORY B00Y MANDS FEET 2 FILM BADGE /TLD C NONE C NONE C NONE C NONE
- MULTIPLEILOCATiON
- HALF MASK C NO PERSONAL C CANVAS GLOVES C SHOE COVERS I
- FULL FACE MASK OLITIR CLOTHING C WATERPROOF GLOVES
- SHOE RUBBERS g GAMMA 00SIMETER SUPPLIED AIR HALF MASK COVERALLS
- DOUBLE WATERPROOF BOOTS RANGE SUPPLIED AIR FULL FACE MASK 2 DOUBLE
- WATERPROOF GAUNTLETS
- NEUTRON BADGE SUPPLIED AIR HOOD
- WATERPROOF GLOVE LINERS HEAD NONE f
- SUPPLIE0 AIR WHEN WEL0tNG.
WATERPROOF HOOD FLAME CUTTING, GRINDING OR
- BETA EYE PROTECTION
- ELECTRONIC 00$lMETER HEATING CONT AMINATED MATERIAL RCT IN ATTENDANCE.
C CONTINUAL TIMEKEEPER REQUIRED C INTERMITTENT INDIVIOUAL RESPONSIBLE FOR OWN TIM! KEEPING mrem / DAY WHILE WORKING UNDER THE PRO PERSONNEL ARE NOT TO EXCfE0 A WHOLE B00Y DOSE OF 05 THIS RWP UNLESS OTHERWISE AUTHORIZED BY RA0tAtl0N CHEMISTRY SUPERVISION SPECIAL INSTRUCTIONS CONTACT RA0 0FFICE FOR MONITORING WHEN RCT:
DATE:
OPERATING SHtFT SUPERVISO'i APPROVAL RAD CHEM. FOREMAN APPROVAL PERIODIC REVIEW REQUIRE 0?
NAME-DATE-YES FREQUENCY:
kLARA ACTION REVIEW (IF RE00iREO)
CNR NO REASON:
NAME:
DATE:
NAME:
0 ATE:
CATE:
7ERMINATION NAME:
REASON. C CONDITIONS CHANGED C JOB COMPLETE JOB CANCELLED C EXPtRED 17a Ci CO e625a2t$n45
RWP INSTRUCTIONS
'" REQUESTING DEPARTMENT Ente
- work desc9pton H es sopede as sp4ce permrts 2 Er,te pt, ma.heur estimate This 4 me best estr. ate of the tota' numbe' of hou's ar pe sonne' *d! spend m the radiation field pederaia; far tast sta*':
on the RWP 3 Eriie work loca9en' Be specific (such as noor, room ecuipment. etc )
l 4 Er'ter the app opnate too catepoay code f om the ist befow l
1 REACTOR OPERATIONS AND SURVI!LLANCES 2 ROUTINE MAINTENANCE 3 IN SERVICE INSPECTI 5 REFUELING 6. WASTE PROCESSWG 5 Enter work recuest tiura.ber and equipment part number,it applicable 6 Check the appropnate t'rork funchon box 7 Enter support group (s) (Escept for Type i Rouhne Access RWP s) This is any grot p needed to mspect or veq the task stated O C. manufactu*ers rep. Tech Staff. RCT) Do not me;ude other worti groups assigned to ths task Suppo'1 pe' sonne' may be ad notificaten of Radiaton Chemistry Supervision 8 Pra recuestei name (must be sornecne knowledgesom m the wo a stated on the RWF; depa tment phone a and cate
" RAD. CHEM. TECH.
1 Enter station number and RWP type.
2 Enter RWP number once the RWP has been approved by Rad:ston Chemstry 3 Attach ALARA Check Lat to me RWP.
4 Attach 131451 survey to the RWP. (ms a required for Type 2 RWP's r/ tty) 5 Check type of survey box. (" PRE J06...", " PRIOR TO. " or "AT START OF,. ")
" PRE. JOB $URVEY" A The t>or will t>e checked when an RCT s gomg to do a pre pb survey for me RWP.
B The RCT will determine the protective equipment requirements from the latest available survey C The RCT wih review the survey data and protectve expment reau.rements with any md'viduats accompanying him on the survey D Page 2 of the RWP will be used for personnel s:gmtures and authonaahon to perform the pre pb survey E Upon completion of the prejob survey, the remamcer of the RWP win be fined out and a new page 2 ntated
" SURVEY PRIOR TO J06" This box wW tie checked when the RCT fills out me RWP based on mformaton from a previous, valid survey
" SURVEY AT START OF J08.
THE USE OF THl$ BOX 15 AN EXCEPTION. ased for jobs whe e personnel safety e nvolved. or wtere tmeiy schon could prevent a unit trip ecoipment damage or the sign.ficant spread of contammaton 6 Enter the work area survey re ds. pnot to the start of the pb. if possible THE SURVEY RESULTS SHOULO INCLUDE:
A Specific dose rates that rmy be o? mierest to the worker, B The contammatcn leve!s (in dothr100cmt) on the souipment, floor and watts of the work. areas indicated C The entremity dose rate and the Estmated Whole Body Dose rate (in mrem mrad /hrt j
- 0. The airborne radcactnnty in me wort area expressed m uCucc or rnpc fraction.
T indicate af temporary shelang is instaned.
l 8 Enter pb man. rem estmate Estimated whole Body Dose rate tmes Man hout Estmate.
j NOTE In$cate "ALAP.A ACTION REVIEW REQUtRE0". as required by station procedures 9 For " CONTACT RAD. OFFICE W51EN...',' flR m any required hold points for Rad.at on Chemistry coverage, if applicable 10 Check proper protectve equ:pment riovired for ths pt.
11 Ind.cate type of bmekeeping and RCT attendance required.
12 Fill m ma6 mum daily dose anowed TYPE 1-50 mrem, TYPE 2-100 mrem.
13 in the spec.a! instructions section, list any informaton necessary for the safe and propt.r performance of mis task.
1a Sign and date the RWP. This is the RCT who fins in the above mformaten.
" RADIATION CHEMISTRY FOREMAN The Ra$aton Chemistry Foreman sha!' review'the ALARA Check List and RWP and sign and date the " Rad Chem Appic.'a:" secton He sha" a'sc review and ap;tte att actve RWP's and s.gn and date weekly on pa:e'2 (Type 1 RWP Approva' Log) or mMat shi*t:y on pa;e 2 (Type 2 RWP App' ova' and S; aturt Log) as required.
" ALARA COORDINATOR Tf? ALARA Coord naiof or h!s de!'gfatsd a'temate shaR a'*ach a written ALARA Acto 7 Revtw and sigi aad date the "A' ARA Ar. ten Reuev." sic'c1 as re-luired by staton procedure.
"0PERATING SUPERVISION Tre s'.*: Eng reer. Operatn; UNt Shi*: Supe visor or Rad. Waste Foreman (for Pad Waste work) seaf' make a "Per.ed.c Reve s." dete".ther a " 4cate 1*e "Fre:Le9ey" of rtview, or state the "Reasci" It a peac$c revtew s not re;uwed He sbasiga :Pd date the "Operata; App
- oval" sect on He ea" also to a'"' dait et page 2 (Type 1 RWP Approia' Lc;) or nta' cn page 2 (Type 2 RWP Approva' ar.d S; nature Lc;) as "Reret. Frecuency" ret. res "TGMINATION
--mun - wwnnah,uex-nnsmnmaamrowe m 4 a.n
- m. that.., tan He c't ngo.ran m the creer scLe v.t: s t at : cWP has
TYPE 1 RUP SIGNATURE LOG RWPA WORK GROUF-PAGC#
i PERSONNEL AUTHORI7ED TO PERFORM WORK I HAVE READ, UNDERSTAND AND WILL COMPLY WITH ALL THE INSTRUCTIONS OF THIS RWP.
Film /TsD Film /TLD Signature Badge #
Date Signature Badge #
Date j
I j
w N*s
.must M
tumumsp 6 m e
l em M
W as NN M
F M
Map
- gammas SED M
l l
AGE #
,dKLY RADIATION-CHEMISTRY FOREMAN REVIEW l
Signature Date Signature Date 1
-l I
A duplicate of this form will be used by Operating Supervision if 1ote:
a periodic re/iew of a Type 1 RWP is required.
6 i
TYPE 2 FWP APPROVAL AND SIGNATURE LOG RUPa WORK GROUP:
ffTIVATION DATE:
TIME:
Init.
tatt.
Init.
Init.
Init.
Init.
Init.
.D.
CHEM.
) REMAN 1
.ST SHIFT
- ND SHIFT e_
3RD SHIFT l
SPROVALS N.R.
Review Frequency:
inst.
Init.
Intt.
- Init, init.
Init.
Init. !
iIFT JPERVISION 1ST SHIFT j
2ND :P11FT e
u._
3RD SHIFT OB FOREMAN init.
Init.
Init.
Init.
Init.
Init.
Init.
ignature i
PERSONNEL AUTHORIZED TO PERFORM WORK
?
I HAVE READ, UNDERSTAND AND WILL COMPLY WITH ALL THE INSTURCTIONS OF THIS RW Film /TI.D Ap rv c,.
Aprvd.
Aprvd.
Aprvd.
Aprvd.
Apevd.
Apryd.
Badge D1y.
Oly.
Oly.
Oly.
Oly.
D1y.
Oly.
f Dose Init. Dose Init. Dose Init. Dose Init. Dose Init. Dose Init. Dose Init
.ignature edueessakun mAEr Wmmunup N
amusuum am a
M M4 wm em N
enmue5ED WE M M
em 6
W46 eM (GM mammens SWN M
Wuummme N*
m eM M
N N
duuuSED M
M GB. MinuMW m
W_
M e
spensium W
emmimus m
I February, 1982 l
1 RADIATION OCCURRENCE REPORTS J
l Purpose j
The purpose of this standard is tot 1.
Describe the circumstances under which a Radiation Occurrence Report
)
should be written; 2.
Provide a means to determine the causes leading to a Radiation Occurrence Report and reduce the likelihood of similar future occurrences; 3.
Provide a means of feedback to manag ement. on the ef f ectiveness of this phase of the ALARA program.
General r
1.
All radiation workers are encouraged to prepare a Radiation Occurrence Report whenever any occurrence, as outlined below; comes to their attention.
q j
NCTIE ; Gcod judgement should be exercised in deciding whether or not the j
it.itiation of a Radiation Occurrence Report is warranted.
In many
]
caCac, the situation may be resolved through less formal channels.
Any violation of technical specifications or regulations ha,vang to do with radiation safety must be written up as a Rediation Occurrence Report.
2.
Once the Radiation Occurrence Report has been initiated, it shall be serially numbered by Radiation-Chemistry and shall be forwardea to the Raciation-Chemistry Supervisor for further consideration.
3.
The Radiation-Chemistry Supervisor or his designated Radiation-Chemistry alternate, shall perform a " follow-up" investigation of the circumstances a rrounding the occurrence in order to deter 6nine the causes and take the appropriate eteps to reduce the likelihood of similar occurrences in the future.
Page 18 of 69
February, 1982 RADI ATION OCCURRENCE REPORTS (CONTINUED)_
4.
If the Radiation-Chemistry Supervisor, or his designated Radiation-Chemistry alternate, believes that a violation of regulatory requirements or a significant violation of the CECO. Radiation Protection Standards has occurred, he shall forward the Radiation Occurrence. Report along with his follow-up to the appropriate station management personnel for f ur tt.ee consideration.
Station management personnel shall act promptly on each Radiation Occurrence Report.
(Follow-up actions should be noted either on the reverse sice of the Radiation Occurrence Report or on an attached sheet).
5.
In reviewing the Radiation Occurrence Report, station management personnel shall:
Initiate action to prevent a recurrence, if practical, and perform the a.
required notifications.
b.
Note any actions taken on the reverse side of the form or on an attached sheet.
Sign the Radiation Occurrence Report.
~
c.
d.
Return the completed Radiation Occurrence Report to the Rad / Chem Supervisor,for filing.
6.
If desired, an employe who initiates a Radiation Occurrence Report may contact management to determine what action was taken on the report.
Types of Occurrences 1.
The following are examples of radiation occurrences relating to the control of external dose equivalent, Improperly, posted or controlled radiation area or high radiation area.
a.
b.
Discovery of working dose equivalent rates to a worker inside a controlled area that significantly exceed the expected dose rates.
c.
Any personnel external dose equivalents in excess of the 10 CFR 20 limits.
i Page 19 of 69
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February, 1982 RADI ATION OCCURRENCE REPORTS (CONTINUED) d.
Any in-plant personnel whole body dose equivalent that exceeds the dose equivalent authorized by Radiation-Chemistry or the administrative controls established in these standards.
2.
The following are examples of radiation occurrences relating to the control of internal' dose and surface contamination:
a.
Skin or personal clothing contamination greater than 1 mR/hr above background as measured with an approved beta-g4mma instrument.
b.
Exposure of personnel to concentrations of airborne radioactive material resulting in an internal deposition in excess of regulatory requirements.
Any case of positive nasal surveys in excess of 10,000 dpm.
c.
d.
Spread of significant contamination in' the immediate work area beyond that which was planned or might normally be expected.
Spread of significant contamination outside of a controlled area.
e.
examp'les of radiation occurrences that relate to the 3.
The following are administrative aspects of radiation protection.
a.
Personnel in a controlled area without adequate oh required approval, timekeeping, monitoring, or dosimetry.
b.
Personnel in a controlled area without required training.
c.
Personnel in a controlled area without adequate protective equipment, d.
Personne"1 in a controlled area improperly using required protective equipment.
Personnel practices in a controlled area in violation of'the Radiation e.
P rotect. ion Standards, Radiation Protection Procedures, facility Technical Specifications, or other regulatory requirements.
j I
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Page 20 of 69
i February,~1982-RADIATION OCCURRENCE REPORTS (CONTINUED) 4.
The following categories are provided for. those situations' or occurrences that should be. reported but do not fit into ' the classifications listed above:
a.
It is believec that a violation of regulatory requirements pertaining to the protection of individuals against radiation has occurred.
b.
A significant action or situation-. inconsistent with the as low as is reasonably achievable (ALARA) philosophy is observed.
c.
Other (specify).
Page 21 of 69
RlDIATION OCCURFOCE REPOC R.O.R. NC.:
Dat e -
- Time:
Type of Occurrence:
(See Reverse)
PERSONNEL INVOLVED SSN TILM BADGE /TLD#
COF2A';Y NAME
'pcation:
l Description of Occurrence (Include all Significant Infors,ation):
Reported By:
Action Taken at Time of Occurrence:
OCCUPENCE PIPOMED TO:
DATE/T1 TITLE
- NAME__,
~
Estimate of Exposure Received:
External:
Internal:
References (WP No., $urvey Report, etc.):
i aa G
Tfrt! Or Oce"estw!!!
Ae *r.!stenttve fert-elt Estearti Dese cert--1 1
Impreperly posted cr contrc!!ee ractation area cr 10 Personnel tr. a cer,tre!!ec area without scer.uatt or recuirec timenerpard, monticrirg. er scriretry.
hign radiation area, 2 Discovery of worning eese rates to a werner Instae 11 Persennel in a centrclite tres althout recuires a contro11e9 area tnat >1gnificantly escred the
- training, espectec cose rate.
Perser.nel in a centrcliec area without scecuste 12.
3 ArJ persennel estert.a! cese ecu1 valent tn escess protective eculpment.
of the 10 CFR 20 limits.
13 Persennel in a centrclied tres im;teretly using 4.
ArJ in-plant persennel whcle bocy ecse eculvalent recuireo trctective eculpment.
that esceecs the cose authcrised Dy Raatation=
Cher.1stry cr the aca,1nistrattve centrcis 14.
Perscr.ne! practice in a contro!!ed area 3r.
estab!!shed in the Reclation Prctection $tancarcs.
Victatten cf the Raelation Prctecticn Stanchr:s.
Radictica Prctection Procecures fac!1*ty Internal Deve are *urf ace Onetamirat t e Centre!
'fechnical Specifications. cr other regu;atcry recuiremer.ts.
5 Sain or personal clothing centamir.atica greater than 1 mR/hr above background as measurec with monettegerstee an approved beta-sama instrument.
- 15. It is believen that a victation cf regulatcry 6.
Drcsurt of perscnnel to concentratter.c of alrDerne requirements pertstr.ths to the prctection cf racicactive material result 1r.g in an internal Individ.ials against ractation has pecurred.
sepcsttlon in escess of regulatory reculrements.
it. A significant attica or situat!cn incenstster.t 7.
ArJ cast of pesstive nasal surveys in tscess of with the A1aPA ph!!csephy is ctserved.
10.000 epr.
- 17. Cther (specify).
3 Strena of significar.t contamination in the twe01ste scra area beycnc that watch was planned cr might hermally De espected.
9.
Spress of significant contestation cutsace cf a centrollte area, chadiation-Chemistry Follow-Up:
Date:
Signes:
nacoltlonal rollow-Up Recosetendea:
(res)
(nc) aseettlonal Pb11c=-Up:
Date:
31gned:
o 9
)
February, 1982 q
- PERSONNEL OCCUPATIONAL EXTERNAL DOSE EQUIVALENT LIMITS This t,tandard Aummarizes current Nuclear Regulatory Comm?ssion (NRC),
- regulations and defines the Commonwealth Edison Company (Ceco.) administrative guide?.ines regarding external occupational doses to radiation workers.
1.
The NRC requires that the following ' occupational external dose equivalent limits not be exceeded:-
]
Part of Body Quarterly Dose (Rems)
Cumulative Dose (Rems) whole body, head and, 1.25*
5(N** - 18) trunk active blood-1 forming organs, lens of eyes, or gonads i
skin of whole body 7.5 NA Hands and forearms, feet 18.75 NA and ankles The 1.25 rem quarterly limit may be increased to a maximum of' 3 rems per j
calendar quarter if, before permitting any individual to exceed the 1.25
)
~
rem limit, the licensee has determined the - individual's accumulated occupational dose to the whole body on Form NRC-4, or on a clear.and legible record containing all the information required,in that form, and otherwise complies with the requirement of 10 CFR Part 20; and determined that the dose to the whole body',
when added to the accumulated occupational dose to the whole body, shall.not exceed 5(N-18) rems.
N equals the individual's age in years at his last birthday.
2.
Pe'rsonnel external dose equivalents at Commonwealth Edison Company nuclear-f.
facilities shall be controlled a2 follows:
a.
Each individual must have supervisory approval " bef ore exceeding a daily whole, body dose equivalent of 50 mrom.
(Refer to the Radiation Work Permit instructions.)
b.
Daily whole body dose equivalents in excess of 100 arem or weekly whole body dose equivalents in excess of 300 mrem must be approved by the Radiation-Chemistry Supervisor or by a health physicist designated by him in writing before authorizing department approval.
I Page 24 of 69 6
February, 1982
- PERSONNEL OCCUPATIONAL EXTERNAL DOSE EQUIVALENT LIMITS (CONTINUED) c.
Quarterly whole body dose equivalents in excess of 1250 mrem to bargaining group personnel (including contract bargaining group per sonnel) can be approved by the Administrative and Support Services Assistant Superintendent after discussion between management and bargaining group representatives.
d.
Annual whole body dose equivalents to bargaining group personnel (including contract bargaining group personnel) in excess of 5000 mrem can be approved by the Station Superintendent af ter discussion between management and bargaining group representatives.
Annual whole body dose equivalents will not exceed 7000 mrem except by mutual agreement between management and bargaining grou- ' representatives.
- System Council U-25 has not reached agreement on all of the criteria specified in this section.
Further bargaining group / management discussions will commence upon promulgation of the revised 10 CFR 20, which is currently under consideration by the PRC, or by April 15, 1983, whichever date is earlier.
Page 25 of 69
February, 1982 PERSONNEL EXPOSURES UNDER EMERGENCY CONDITIONS Purpose J
This standard provides recommended guidelines for the control of voluntary personnel exposutes under emergency conditions.
'l Recommended Limits The dose-limiting recommendations for emergency situations are as followu l
1.
Emergency Dose Limits - Life Saving Actions:
(This applies to search and removal of injured persons or entry to the emergency area to prevent conditions that would injure other people).
Dose equivalent to the whole body shall not exceed 75 rem; a.
b.
Additional dose equivalent to the extremities shall not exceed 200 rems.
2.
Emergency Dose Limits - Less Urgent Actions (This applies under less stressful circumstances where it is still desirable to enter the emergency area to protect facilities, eliminate serious unplanned release of effluents, or to control fires),
a.
Planned whole body dose equivalent shall-not exceed 25 rems; b.
Planned dose equivalent of the extremities shall not exceed 100 rems (including the whole body component).
c.
Planned dose equivalent to the thyroid shall not exceed 125 rems.
Requirements 1.
The emergency limits outlined above shall be voluntary and shall not be used except in case of a bonafide emergency involving risk of life or limb, or the destruction of valuable preparty; normally, an acute whole 1
body dose equivalent in excess of 25 rem shall be limited to once in a-lifetime.
l i
i Page 26 of 69
r-Fe brua ry, 1982 PERSONNEL EXPOSURES UNDER EMERGENCY CONDITIONS (CONTINUED) 2.
Whenever possible, the prior approval of the Station Superintendent, CECO.
Medical Director, and the Radiation-Chemistry Supervisor shall be obtained before exposing personnel to dose equivalents in the ranges specified
- above, and in
- addition, emergency personnel shall te informed before-the-fact of possible health-ef fects at the anticipated exposure levels.
3.
Whenever possible, respiratory protection and protective clothing shall be worn to reduce internal and external contamination to levels comparable to those encountered under normal operating conditions.
4.
Women should not be assigned work involving emergency exposures unless approveo by the CECO. Medical Director and the individual involved.
Approval shall be obtained before the fact.
5.
The CECO. Medical Director shall be promptly notified of any internal or external dose equivalent which exceed a regulatory limit.
6.
Personnel uose equivalent resulting f rom an emergency situation shall be estimated and reccrded.
- 7. - Personnel internal or external dose equivalent in excess of the regulatory' limits shall be investigated by station management and reported to the NRC
]
and to the individuals as required by 10 CFR Parts 19 and 20.
l Page 47 of 6*
n 1
February, 1982 REMOVAL'AND RETURN OF PERSONNEL FROM RADIATION WORK.
.]
Purpose This standard outlines those conditions under which personnel will not - be allowed to work in controlled areas, and describes - the cotiditions which' must be satisfied before an individual may be returned to work.
l Restrictions 1.
In the case of known exposure to radiation in excess of the quarterly, or accumulated lifetime limits specified in 10 CFR Part 20, the individual I
involved shall not be permitted to enter controlled areas (until the possibility of exceeding any of ' the remaining limits no longer exists).
Return to work in a controlled area is contingent.upon the careful evaluation and regalation of the individual's work status.
NOTE:. The CECO. Medical Director shall be promptly notified of any internal or external doses which exceed a regulatory limit.
2.
In the case of a suspected exposure to external radiation in excess of the j
quarterly or accumulated lifetime limits:
a.
The individual shall not be permitted. to enter controlled areas (except upon the recommendation of the Radiation-Chemistry Supervisor) entil an evaluation of dosimetry results indicate whether or not an exposure in excess of the applicable limits has occurred:
b.
If an exposure in excess of the applicable limits has occurred, the i
instructions in section 1 above apply if the exposure has not exceeded these limits, the individual may be returned to raalation work.
3.
In the event that any of the permissible external dose equivalent. limits are being approached and further exposure might reasopably be expected to result in exceeding either the quarterly or accumulated lifetime limits, the individual involve <3 shall not be perraitted to enter controlled areas until such time as the possibility of exceeding the permissible limit no longer exists.
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Page 28 of 69
)
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l 1
February, 1982 REMOVAL AND RETURN OF PERSONNEL FROM RADIATION WORK 4.
In those cases where an individual is involved in an occurrence and a significant body burden is possible or in the event that a bioassay result exceeds 104 MPOB for any nuclide or organ or.when the whole ~ body bioassay result exceeds 25% MPOB for the sum of all organs.
a.
The individual involved shall not be permitted to enter " airborne radioactivity" areas or " radioactive materials" areas which contain smearable contamination until such time as the bioassay. results have been evaluated.
b.
Personnel with body burdens greater than the permissible limit shall
'i not return to radiation work until approval is_ granted by the CECO.^
Medical Director and the Radiation-Chemistry Supervisor.
l Page 29 of 69
February, 1982 DOSE EQUIVALENT RATE DETERMINATION i
Purpose This standard describes practices which shall be observed when using. survey J
instruments to establish dose equivalent rates.
I General 1
1.
Contact Radiation-Chemistry for specific instructions concerning the cperation and interpretation of survey instrument responses.
f 2.
All dose equivalent rates shall be determined by Radiation-Chemistry.
)
3.
Any individual who has received Nuclear Station General Employee Training may use a survey instrument to verify working dose rates which have been determined by a Radiation-Chemistry Technician.
4.
.Si nc e very large and sometimes unknown correction factors are involved when attempting to determine suzf ace dose equivalent rates of very small sources, direct contact with thim mu'st be avoided whenever possible.
Monitoring
.o 1.
Beta-gamma survey instrume'nts* shall be checked for proper response prior to the start of a job and at I. east every four hours while in use.
If an erratic or other unusual response is noted, or if for any reason the user doubts the accuracy of the inytrument, a substitute instrument shall be obtained immediately, and Radiation-Chemistry should be notified.
2.
When a GM-type instrument is used in lieu of an ion chamber instrument to determine personnel dose equivalent' rates the instrument must incorporate energy compensated /linearization circuitry over the range of energies and dose equivalent rates expected.
The operator shall be aware of general saturation limitations of a GM type instrument.
3.
Earphones or speakers should be used with GM-type instruments when making contamination surveys if the instrument has speakers or earphones.
(
e Page 30 of 69
February, 1982 DOSE RATE DETERMINATION (CONTINUED) 4.
Contamination control bagging must not exceed 10 mg/cm2 in thickness over the beta window portion of dose rate monitoring instruments without Radiation-Chemistry _ approval.
5.
survey instruments with glass or similarly fragile detecting tubes must not be used for surveys of facial areas.
Shielding Considerations 1.
The shielding afforded by' masks and face shields.may be routinely considered in establishing timekeeping and whole body dose equivalent rates to the eyes provioed the eyes are effectively, shielded from beta radiation from all angles.
2.
The shielding af forded by gloves and boots may be routinely' considered in t
establishing skin and extremity beta dose equivalent rates used for establishing dose equivalents.
3.
The shielding effects of clothing for other parts of the body may be considered in establishing timekeeping beta dose equivalent-rates only upon the specific authorization of the Radiatiort-Chemistry Supervisor or the Lead Health Physicist.
i Page 31 of 69
February, 1962 PERSONNEL DOSIMETERS Purpose This standard provides guidelines for the appropriate use of personnel dosimeters in order to comply with regulatory requirements and with Ceco, practices.
General 1.
Personnel monitoring equipment, as noted in this standard, refers to all devices designed to be worn or carried by an individual for the purpose of measuring the dose equivalent received (e.g.,
film badges, TLDs, pocket dosimeters, electronic dosimeters, etc.).
2.
Each person shall wear all personnel dosimeters as prescribed by Radiation-Chemistry. Dosimeters will be stored at locations designated by Radiation-Chemistry when not in use.
3.
Personnel monitoring equipm.ent employed to meet the requirements of this standard must be approved by the Radiation-Chemistry Supervisor.
4.
When personnel dosimeters have been lost or damaged, as for example by excessive
- heat, moisture,
- droppage, or severe vibration, or when dosimeters have been contaminated, the person involved shall l' eave the controlled area and notify his immediate supervisor and Radiation-Chemistry.
Regulatory Requirements NRC regulation 10 CFR Part 20 requires each licensee to supply appropriate personnel monitoring equipment and to require the use of such equipment by:
1.
Each individual who enters a controlled area under such circumstances that he receives, or is likely to receive, a dose equivalent in any calendar quarter in excess of 25 percent of the applicable quarterly limit.
2.
Each individual under 18 years of age who enters a controlled aren under circumstances that he receives, or is likely to receive, a dose equivaler,t in any calendar quarter in excess of 5 percent of the applicable quarterly limit.
3.
Each individual who enters a high radiation area.
Page 32 of 69
3
-1 February, 1982 PERSONNEL DOSIMETERS (CONTINUED)
J
)
Wearing Personnel Dosimeters Wearing of personnel dosimeters, when required by the regulatory requirements 1
listed above, will be subject to the following guidelines:
1.
Normally, the personnel dosimeters shall be, worn near each other on tne f ront part of the body at or above the waist level.
2.
On those jobs where the most restrictive dose equivalent is to the head, back, upper arms, lens of eyes, gonads, or legs, Radiation-Chemistry i
personnel will. require that whole body dosimeters be worn in the region of the highest whole body dose rate.
3.
Additional dosimeters shall be issued to supplement whole body dosimeters when the limiting dose equivalent rate is to the extremities.
4.
Ionization chambers shall not be worn clipped together:
When two ionization chambers are used, one is worn on each side of the film badge /TLD.
j 5.
Care should be taken to prevent contamination of dosimsters.
The f
following materials, in order of preference, are recommended when it is necessary to completely cover the film badge /TLD and ionization chambers j
to prevent them f rom becoming contaminated:
a.
Small approved plastic bagt use only one layer of plastic.
1 b.
One layer of masking tape.
c.
One layer of non-waterproof RWP clothing.
6.
It may be necessary to take credit for beta radiation shielding afforded by wearing rubber or similarly effective protective apparel' which completely covers that portion of the body exposed to 'the beta radiation, j
Under these circumstances all dosimeters are worn under the protective apparel.
7.
An individual whose self-reading dosimeter reads off-scale shall leave the work area and report to his rupervisor and then to Radiation-Chemistry immediately.
i s
Page 33 of 69 e
February, 1962 PERSONNEL DOSIMETERS (CONTINUED) 8.
On jobs where self-reading dosimeters are issued, each individual shall read his dosimeter periodically in order to be aware of his accrued dose as the job progresses, visitors The guidelines for use of personnel dosimeters by f acility visitors are as follows:
1.
Ncnstation personnel, on-site for less than one day normally are not i
cadged when visiting or working in uncontrolled areas.
2.
Visitors normally are not badged when escorted by to'ur guides through uncontrolled areas.
3.
When none of the conditions listed in the " Regulatory Requirements" l
I section of this standard apply, tour guides who escort visitors through controlled areas shall carry a group badge and extra ionization chambers and issue an ionization chamber to each member of the tour group on each tour so that all. individuals in the group may be collectively monitored.
The requirement to provide an ionization chamber to each member of a tour group may be waived if the tour has specific approval of the Station Superintendent.
4.
When any of the conditions listed in the " Regulatory Requirements" section of this standard apply and tour guides are not used to escort the visitors through controlled areas, each visitor shall be provided with appropriate personnel monitoring equipment and shall be required to use such equipment.
Page 34 of 69
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February, 1962 j
l OCCUPATIONAL RAPIATION EXPOSURE RECEIVED AT OFI-SITE LOCATIONS i
Purpose This standard describes Company practices concerning the monitoring of occupational radiation exposure received at off-site l aations.
General 1.
While visiting off-site locations where exposure to radiation or radioactive materials is involved, each individual should make every effort to maintain his exposure ALARA and whenever possible should ascertain his quarterly, and cumulative dose before leaving.
2.
Personnel visiting any off-site location shall immediately inform the Radiation-Chemistry Supervisor at his assigned station of any exposure received off-site in excess of the limits specified in 10 CFR Part 20.
Occupational Exposure Received at Other CECO. Facilities Personnel visiting' other CECO. facilities where exposure to radiation or radioactive materials is involved will wear a film badge /TLD and any other monitoring devices as issued by that facility's Radiation-Chemistry personnel; personnel dose received at the site will be transferred to the appropriate facility records.
Occupational Exposure Received at Non-CECO. Facilities j
1.
Personnel visiting non-CECO. facilities where exposure to radiation or radioactive materials is involved will account for dose equivalent received as follows:
a.
Use the film badge /TLD and any other monitoring devices as issued by that facility's Radiation-Chemistry personnel.
Request a written exposure estimate upon completion of the assignment at the facility prior to leaving.
Upon return to a Ceco. facility the individual j
shall inform the Radiation-Chemistry Supervisor of any exposure i
f received off-site.
a b.
Prior to visiting the non-CECO.
facility notify the Radiation-Chemistry Supervisor for any special instructions concerning use of regularly assigned Ceco. dosimetry.
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Page 35 of 69 f
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OCCUPATIONAL RADIATION EXPOSURE RECEIVED AT OFF-SITE LOCATIONS (CONTINUED) l 2.
All personnel who receive occupational exposure at off-site locations shall provide information regarding whom to contact for a complete exposure record for the period involved if such information' is not provided when they leave the f acility.
Also the individual must provide bioassay samples as may be requested by the Radiation-Chemistry Supervisor.
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Pa9e 36 of 69
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February, 1982 1
RESPIRATORY PROTECTION j
Purpose This standard describes those requirements that are to be observed when using respiratory equipment for radiological respiratory protective purposes.
General 1.
Respiratory protective devices employed to limit concentrations ~of l
radioactive materials in air in restricted areas to acceptable levels must l
be used in compliance with all applicable requirements. stipulated i n. 10 '-
CFR 20..
2.
Only respiratory protective equipment which has been approved by Radiation-Chemistry supervision shall be worn for radiological respiratory-protective purposes.
3.
Prior to' performing task's requiring the use of respiratory protective equipment, each individual and his immediate supervisor must' have received adequate tra'ining on each type of respirator to be used as required by-10 CFR 20.
Each individual must have medical l pproval prior to-being assigned work requiring the use of a respirator.
"4.
Individuals using respirators which require a
facir.1 seal for ef f ectiveness must not have any f acial hairs, or wear protective apparel or devices (e.g. glasses with bows, hearing aids
...)
which interferes with the sealing surf ace of the respirator.
5.
Contact lenses must not be used by individuals wearing f ull'f ace supplied j
air, full face cartridge respirators, or supplied air hoods.
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6.
All maintenance of respiratory protective equipment must be approved by j
Radiation-Chemistry supervision.
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7.
Radiation-Chemistry personnel will prescribe the appropriate respiratory protective equipment to be worn for each job, based, on 10 CFR-20 requirements.
1 8.
Each individual must inspect-and ensure operability of respiratory protective equipment prior to entering the ' controlled area.
Page 37 of 49 i
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February, 1982 1
RESPIRATORY PROTECTION (CONTINUED) 9.
An individual may leave an area at, any time ~ for relief from respirator use
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in the event of equipment - malf unction, undue physicalor, psychological stress,. procedural or communication f ailut es, significant deterioration of operating conditions or any other reason that.the user feels ~ is necessary for relief.
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- 10. Each mask. brought to the work area must.be. carried in a clean plastic bag and handled in such a manner L as to avoid contaminating the respirator.
After use, respirators must be individually begged and deposited in an~
1 appropriate container.
Assessment of Intake 1.
Documentation 'of an individual's inhalation of airborne radioactivity in terms of MPC hours shall be maintained in accordance with 10 CFR 20.
2.
If modes of intake other than inhalation of airborne radioactive materials are present, they must also be considered ;in the. dose evaluation when protective equipment for these other pathways is not practical or is otherwise unavailable.
3.
The need for bioassays to evaluate an individual's internal exposure will be determined by Radiation-Chemistry supervision.
4.
Records of bioasssays, employee training, respirator fit tests, medical 1
qualifications, etc. shall be maintained in accordance with 10 CFR 20.
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February, 1982 CONTROLLED AREA' INJURIES Purpose This standard provides Radiation-Chemistry guidance for:
1 1.
Personnel injured in.a controlled area.
j 2.
Personnel with skin injuries who work.in controlled areas.
4 Controlled Area Injuries Contro11ed' area injuries shall be immediately reported to supervision and to Radiation-Chemistry and the.following actions initiated l
1.
A contamination survey of the injury, the object causing the injury, and any clothing which may have been pierced by the ob3ect inflicting the injury.
If survey results indicate a
contaminated
- wounc, the 1
Radiation-Chemistry Supervisor,shall be notified immediately.
2.
If. the-injury is of a serious nature, contamination control shall be limited to that required to prevent spreading contamination into the, wound.
If.the injury is of a
relatively minor
- nature, normal contamination control ' procedures should be followed as closely as possible.
Skin Breaks 1.
Each individual must notify his supervisor and Radiation-Chemistry of any skin breaks or skin conditions which.he may have prior to performing ~ work i
in a contaminated area.
The supervisor shall insure that all skin _
injuries or skin conditions are examined by Radiation-Chemistry and adequately protected before permitting the worker to enter a contaminated area.
Skin inju' ries or skin conditions shall include unhealed wounds, open cracks Irom chapping, and other injuries such as punctures, abrasions, lacerations and incised wounds.
2.
The individual should understand that approval-to ' perform work in a contaminated area is contingent upon. the protection af forded the wound. and that he should make every reasonable effort to maintain the dressing and-bandage in a satisfactory condition.
Page 39 of 69
February, 1962 PERSONNEL BIOASSAY SAMPLING FREQUENCY Purpose This standard provides bioassay sampling frequencies for all workers at CECO.
nuclear facilities.
General 1.
Bioassays (in-vivo measurements and/or measurements of radioactive material in excreta) will be conducted as necessary to aid. in determining the extent of an individual's internal exposure to concentrations of l
radioactive material.
2.
Exceptions to the bioassay sampling frequencies described below may be made by the Radiation-Chemistry Supervisor if survey da tat indicate such j
analyses are unnecessary.
If an exception is made it should be recorded along with the reason for the exception.
3.
Any individual or work group may be bioassayed more frequently than required below upon the authorization of the Radiation-Chemistry Supervisor.
4.
Recording and reporting of bioassays, as specified in 10 CFR 19 and 20:
a.
Records of bioassays chall be preserved until the station is authorized to dispose of 'them.
b.
When an individual terminates his work assignment at a nuclear f acility, a record of bicassays performed must be provided to the NRC and to the individual.
Bicassay Sampling Frequencies Minimum bicassay sampling frequencies of permanently as:ligned station personnel and long term contractor personnel at operating nuclear stations will be as follows:
1.
Personnel listed in this category are not normally required or expected-to work in radioactive materials areas and/or airborne radioactivity areas and will be bioassayed at least once per year.
Storeroom, Clerical, Janitors, and Security Perscnnel a.
Page 42 of 69
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February, 1982 WELDING, FLAME CUTTING, GRINDING OR REATING RADIOACTIVE MATERIALS Purpose This standard provides guidance for the use of respiratory protective equipment when welding, flame cutting, g rinding or heating radioactive materials.
General l
4 1.
Only those respiratory devices approved by Radiation-Chemistry may be useo in the protection of personnel when welding', flame cutting, g rinding or heating radioactive materials.
j 2.
The use of approved respiratory devices is subject to the qualifications
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outlined in the Rad!: tion Protection Standards, " Respiratory Protection."
Respiratory Protection Requirements I
1.
Unless exempted in a* written, posted notice by the Radiation-Chemistry-j Supervisor, positive pressure supplied air respiratory equipment shall be used when welding, flame cutting, grinding or heating materials which are known or suspected to be radioactive.
2.
When welding, flame cutting, grinding or heating materials which are suspected of being radioactive, the need for respiratory protection should be evaluated by Radiation Protection.
3.
Other operations with radioactive materials (e.g. high speed sawing) which' may involve high temperatures and possible volatilization of the material or may cause radioactive particulate to become airborne should be evaluated on a case by case basis by Radiation Protection to determine the need for respiratory protection.
Deviations Deviations f ron this standard may be authorized by a written posted notice at the job site and at the Radiation-Chemistry of fice which shall be signed by the Radiation.themistry Supervisor.
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Page 41 of 69 j
o February, 1982 CONTROLLED AKEA INJURIES (CONTINUED) 3.
If, at any time during the course.of performing work in a controlled area, it is determined that a skin injury or skin condition is not adequately protected, 'the worker shall' leave the area and notif y Radiation-Chemistry.
immediately.
First Aid Measures 1.
Plant staff should be thoroughly f amiliar with the location of first aid equipment ano equally f amiliar with the methods for use of the protective and first aid equipment which will be used in the treatment of injured personnel.
2.
First aid measures should be administered by plant personnel who have been trained in first aid courses which have been approved by the Ceco. Medical Department.
3.
Instructions on first aid, decontamination and evacuation of exposed and/or contaminated personnel resulting f rom an emergency situation, are provided in the Emergency Plan Implementing Procedures of the Generating Station Emergency Plan.
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Page 40 of 69 i
February, l982 PERSONNEL BIOASSAY SAMPLING FREQUENCY (CONTINUED)-
b.
Contractor and Station Administrative Personnel c.
Training Personnel d.
Nuclear Station Operators e.
Other (personnel normally categorized under 2. or 3. below who do not work in radioactive materials areas. and/or airborne. radioactivity areas.)
2.
Personnel listed in this category may on occasion perform work in areas-l J
contair.ing radioactive materials and/or an airborne radioactivity area and.
will be bioassayed at least twice' per year.
a.
Technical Staff Management b.
Operating Management c.
Contractor Operational Management d.
Electrical and Instrument Maintenance Management" g
I 3.
Personnel listed in this category routinely. perform work in areas
- 1 containing radioactive materials or in airborne radioactivity areas and will be bioassayed at least three times per year.
l a.
Equipment Operators, Equipment Attendants I
b.
Stationmen c.
Mechanical Maintenance Personnel and Management d.
Electrical Maintenance Personnel e.
Instrument Mechanics
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Fuel Handicts & Management g.
Radiation-Chemistry Technicians and Management h.
Contractors - Non Management Page 43 of 69 4
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i; February, 1982
-j PERSONNEL BIOASSAY SAMPLING FREQUENCY.
j (CONTINUED)
Minimum bioassay sampling frequencies of short term non-permanently essigned
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company ana contractor personnel will be as follows:
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1.
All personnel not permanently assigned to the station who work in a radioactive materials area and/or an airborne radioactivity area shall receive a bioassay upon termination of the work assignment at the station.
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l Page 44 of 69
February, 1982 RADIOLOGICAL PROTECTIVE CLOTHING (cot 1TINUED) 2.
All: protective clothing _ that is contaminated between 5 mR/hr and 25 mR/hr at one inch with a calibrated monitoring instrument requires approval from Radiation-Chemistry supervision. before such clothing. is decontaminated in the laundry / dry cleaning facility.
3.
Protective clothing contaminated less than 5 mR/hr ; at 'one inch may be decontaminated in the ' laundry / dry cleaning facility without special
-approval er handling procedures.
4.
Clothing containers should be placed at the exits f rom all' areas where protective clothing is required.
NOTE: Contaminated laundry should not be aj. lowed to 6ccumulate at tne job'
- site, 5.
All protective clothing must be inspected and' decontaminated (if necessary) by 19undry personnel in a CECO. nuclear laundry / dry cleaning facility or in a licensed commercial facility.
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February, 1982 RADIOLOGICAL PROTECTIVE CLOTHING i
Purpose I
This standard provides ' acceptable criteria.for the use and disposition of i
radiological protective clothing.
(Protective-clothing includes all l
radiological protective apparel).
1 Ge ne r a,,,
1.
Only those items of radiological protective clothing wh'ich have been approved by the ' Radiation-Chemistry-Supervisor shall be used 'for protecting personnel from radioactive contamination or beta radiation..
1 2.
The Radiation-Chemistry Supervisor shall be responsible -for providing sufficient quantities of various items of protective clothing in usable condition for work in contaminated areas.
.i.
Radiological protective equipment shall be-stenciled,
- colored, or otherwise marked so as to be readily identifiable.
4.
Personnel must inspect all radiological protective equipment before use to ensure that it will provide adequate contamination protection.- Defective
, protective clothing shall be discarded in a contaminated waste container.
Contact Radiation-Chemistry regarding disposition of all other defective protective equipment.
5.
Unless otherwise authorized by Radiation-Chemistry Supervision, protective clothing shall only be worn in controlled areas.
Radiological protective clothing shall not be substituted for contaminated 6.
ll personal clothing.
Rewearing Radiological Protective Clot h 1.
Protective clothing may be reworn in an area of low-level contamination if approved by Radiation-Chemistry prior to the start of the job and provided t.he wearer is reasonably certain that the clothing is free of loose contamination.
Disposition of Contaminated clothing 1.
Protective clothing contaminated above 25 mR/hr at one inch shall be disposed of as contaminated waste.
NOTE:
Radiation-Chemistry personnel will recommend special handling and segregation procedures for all protective clothing contaminated above S mR/hr.
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4 September 1985 1
f UNCONDITIONAL RELDjjS Purpose This standard'provides guidance for obtaining an unconditional release for materials from a controlled area.
General The removal and re-use of material without restrictions from a 1.
controlled area requires an unconditional release by Each individual removing material from a Radiation-Chemistry.
controlled-area will be responsible for complying with this requirement.
It shall be the responsibility of the. individual involved to request A re-survey shall be 2.
Radiation-Chemistry to make the release survey.
made of the material prior to removal from the controlled area if more than three days have elapsed since the initial survey was made or if-there is reason to suspect that recontamination may have_ occurred.
Radiation-Chemistry may unconditionally release materials by either completing the " unconditional release"-card or by verbal communication 3.
with the requestor, whichever method is more practical at the time of the survey.
When the material released is to remain on-site, unconditional release cards should not be removed until the material is removed from the 4.
When the destination of the material released is controlled area.
off-site, unconditional release cards should not be removed until the material leaves the site.
Personnel approving " Building or Site Passes" must note the following on j
5.
the pass:
Yes No Unconditional Release - Required:
Obtained:
Yes No ReJease Criteria Radiation-Chemistry will determine that all surf aces of the material.are free of any removable or fixed contamination before an unconditional release Surveys must be made for all emitters unless it is known that a is issued.
Material and equipment will be given 4
given type of radiation is not present.
an unconditional release by Radiation-Chemistry provided:
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February, 1982 PERSONAL EFFECTS PERMITTED IN CONTAMINATED AP.EAS j
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Purpose This standard specifies requirements concerning the wearing of personal effects in contaminated areas.
Personal Items which may be Worn in Contaminated Work Areas l
1.
In work areas where "No Personal Outer Clothing" is specified, only those authorized protective and personal equipment items listed below may be wornt a.
Underclothing, shoes, and socks.
i b.
Prosthetic devices and/or glasses, hearing aids, etc., or other items necessary to the safety of the worker entering the area.
2.
In work areas when minimal protective clothing is specified, only the j
following items may be worn:
1 4
i 4.
All authorized protective equipment and personal items listed above, i
1 b.
Slacks, shirt or blouse, and belt or suspenders.
Personal Items not Authorized in Contaminated Work Areas 1.
Watches, rings, cigarette lighters, wallets and contents thereof, fountain pens and penc il.s,
and all other jewelry and personal property not necessary to the safety of the employee in the area.
2.
Suit or sport coat, sweaters, vests, and ties.
Page 47 of 69
September 1985 No contamination above background is detected using; a.
1)
Fixed laboratory counting instrumentation when the geometry and orgin of the material or equipment permit such counting; o_r 2)
Portable survey instrumentation when the use of laboratory counting equipment is not practical. The instrumentation and techniques must be capable of detecting 5000 dpm/100 cm2 total and 1000 dpm/100 cm2 removable beta-gamma contamination, and 100 dpm/100 cm2 fixed and 20.dpm/100 cm2 removable alpha contamination; and b.
There is no likelihood of trapped or occluded radioactive materials in the material or equipment.
gxceptions Radioactive materials shipments are excluded from the requirements described j
herein.
For instructions relating to such shipments refer to the standard-concerning off-site shipment of Radioactive Material.
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February, 1982 TRANSPORTING RADIOACTIVE MATERIALS WITHIN THE RESTRICTED AREA Purpose This standard provides guidelines for the transporting of radioactive materials within and between controlled areas of the restricted area.
General 1.
Contact Radiation-Chemistry for recommended monitoring and protective equipment requirements, if any, prior to start of the job.
Radiation-Chemistry must be inf ormed before the fact if radioactive materials are to be removed from controlled areas.
NOTE: Protective equipment is not normally worn in uncontrolled l
areas.
2.
Preparation of radioactive materials for transport between controlled areas of the restricted area will be conducted in accordance with the followings a.
All materials being transferred must be packaged or contained to the extent that there is no possibility of spreading contamination to uncontrolled areas.
b.
All materials being transferred shall be clearly marked with radiation symbols and should include any additonal information necessary for radiation safety, such as description of contents, activity in curies, date, dose equivalent rates at contact or at specified distances.
3.
When moving radioactive materials within any part of the restricted area:
Minimize personnel dose equivalent whenever practical by:
a.
(1)
Avoiding direct contact with the materials.
(2) The appropriate placement of shielding be tween personnel and materials.
(3)
Limiting the time spent in handling the materials.
(4)
Increasing the distance between the worker and the source.
Page 50 of 69
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February, 1982 TRANSPORTING RADIOACTIVE MATERIALS WITHIN THE RESTRICTED AREA Radiation-Chemistry for monitoring whenever a dose equivalent b.
Contact rate of greater than'25 ares /hr at six inches is* detected.
NOTE: Private vehicles must not be used to. transport radioactive materials within the restricted area.
4.
Radioactive materials should ~ not be lef t unattended while enroute until placed in a properly posted controlled. area.
Upon arrival at the destination the material'shall be placed in a properly 5.
posted controlled area.
NOTE:
Radioactive materials prepared for shipment and packaged' and labeled in accordance with the regulations of the Department of Transportation are exempted from this requirement.
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6.
The control room should be notified prior. to movement of radioactive zaterials which may cause area radiation monitors to alarm.
Spills or Accidents 1.
If a spill or accident occurs while enroute, the ace,ident area should not be left
-unattended until it has bee _n properly controlled.
Radiation-Chemistry shall be notified of the occurrence as soon as possible.
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j February, 1982 l
I STORACE OF CONTAMINATED MATERIALS (CONTIICEDL l
High Level Contaminated Tool Storage Roorn 1.
Contaminated tools and equipment exceeding
- 5. mR/hr above background at shall be stored in a high level contaminated tool storage room or contact area as near the zone of use as prectical.
The exposure rate at the room or area boundary shall not exceed 5 mR/hr.
2.
3.
. Specific authorization must be obtained from the work group supervisor before these tools are transported.
Contact Radiation-Chemistry for an evaluation when the whole body dose equivalent rate to the worker exceeds 25 mrem /hr.
4.
No tools or equipment stored in this zone shall read more than 100 mR/hr at contact-with the wrapping or tool storage container.
5.
Tools removed f rom the tool - room must be packaged as necessary to prevent spreading contamination into the area of transf er. All tools removed from contaminated pools or other significantly contaminated areas must be reasonably decontaminated, wrapped as necessary, and labeled " Radioactive Materials" before transfer to the tool storage room.
Significant dose equivalent rates, date and survey data should be included on the radioactive materials label.
(See note under section 3.
" Low Level Contaminated Tool Storage Room *).
Indoor Storage of Contaminated Waste Materials 1.
All radioactive wastes shall be packaged in containers approved by the Radiation-Chemistry Supervisor.
2.
Containers intended for off-site disposal must be individually identifiable, surveyed, and labeled ' Radioactive-LSA" or as otherwise appropriate and placed in a properly posted and controlled storage area.
Smeurable contamination on the exter:ior surf aces of containers placed in these areas must meet the requirements of the Department of Transportation, the tE, and the appropriate burial sites.
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Outdoor Storage of Contaminated Tools and" Equipment j
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1.
Outdoor s';orage of contaminated tools and equipment shall only be used j
when indcor storage is not'available.
Material placed in outdoor storage will be moved indoors as soon as space becomes available.
l P&ge 53 of 69
February, 1982 STORACE OF CONTAMINATED MATERIALS Purpose This standard pr'ovades guidelines for the storage of contaminated materials.
General 1.
Contact Radiation-Chemistry for protective equipment requirements and survey information when handling contaminated
- tools, equipment, or radioactive materials.
2.
Contaminated tools and equipment from the contaminated tool storage rooms should be identified by magenta or purple paint whenever practical.
3.
The number of storage areas should be limited as practical.
Temporary storage areas should be cleaned up and released when no longer needed.
^
Low Level Contaminated Tool Storage Room 1.
The gamma activity on tools stored in this area must not exceed 5 mR/hr above background at contact.
Each tool must be surveyed and wiped of f or decontaminated as necessary before returning to storage.
2.
The exposure rate at the room or area boundary shall not exceed $ mR/hr.
3.
No specific authorization o'ther than the requirement to perform work in a controlled area shall be required to obtain and transport these tools to the work area.
Tools shall be wrapped, bagged or otherwise contained and labeled radioactive materials before movement between controlled areas.
Note:
Tools may be placed in t.n approved cart for transfer within the' plant. Such tools shall be paintcd to indicate contamination but no label is needed.
4.
Minimum protective equipment for handling tools in these rooms or areas shall be RWP gloves.
Page 52 sf 69
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FGbrusty, 198'2 l
SMRAGE OF CONTAMINATED MATERIALS (CONTINUED) l 2
2.
All contaminated tools and equipment with greater than 2200 dpm/100 cm 2 for alpha emitters in for beta-gamma emitters or 220 dpm/100 cm smearable form shall be enclosed in weather-proof containert. bef ore being -
i stored on-site in outdoor controlled areas.
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3.
Packages must be surveyed and labeled, and periodically checked by the person responsible for the package to ensure that all weather-proofing j
remains intact.
This person shall also arrange for periodic surveys of
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the package.
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Storage of Contaminated Tools and Equipment in Company-Owned Railroad Cars 1.
only totally enclosed and properly posted railroad cars shall be used for i
storing contaminated materials in railroad equipment located on-site.
Unattended cars used for this purpose must be locked when positioned outside controlled areass 2
2.
Tools and equipment contaminated with more than 2200 dpm/100 cm 2 for alpha emitters smearable for beta-gamma emitters or 220 dpm/100. cm must be suitably wrapped before storing in the car.
When these materials are contaminated to less than the levels listed above, they will not require packaging prior to transfer.
3.
Radiation-Chemistry must be advised whenever materials are placed in the car.
Page $4 of 69
February, 1982 OFF-SITE SHIPMENT OF RADI0 ACTIVE MATERIALS Pyspose This standard provides guidelines for off-site. shipment of radioactive materials.
General 1.
This standard does not - apply to. the shipment of materials whose specific activity is less than 0.002 uCi/gm (low specific activity as defined by D.O.T.)
and in. which the radioactivity is essentially uniformly 1
distributed.
2.
Caution should be exercised while packaging or handling radioactive materials to prevent contamination and to minimize personnel exposure.
.I Shipment of Radioactive Material i
1.
All shipments must be properly classified, described, packaged, marked and labeled and be in the proper condition for transport 'for the appropriate 1
mode of transportation in accordance with the applicable regulatory requirements.
2.
All shipments must be properly packaged and labeled in accordancel with requirements set forth by the receiver of the shipment' (i.e.
state
- requirements or burial site requirements).
3.
Prior to shipping radioactive materials off-site, the Radiation-Chemistry Supervisor should verify that the receiver is licensed to receive the shipment.
4.
A
- Radioactive Materials Shipment Record" must be processed for_ each off-site shipment, with a copy maintained at the station.
A copy of tne shipment record must accompany the package in transport.
NOTE: For off-site shipments by aircraft, two copies of the shipment 1
record must accompany each shipment.
j 5.
The Radiation-Chemistry Supervisor or Lead Health Physicist must certify j
that each off-site shipment of radioactive materials meets all applicable i
regulations.
1 Page 55 of 69 1
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February, 1982 RECEIPT OF RADIOACTIVE MATERIAL FROM OFFSITE 1
j Purpose This standard provides guidelines for receiving from offsite locations radioactive material in excess of " Exempt" quantities with the exception of certain isotopes and their associated activity limits as specified in 10 CFR 20.
)
j General
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l.
Pernonnel ordering radioactive materials or equipment incorporating l
I rad ioactive materials must inform the Lead Health Physicist or Radiation-Chemistry Supervisor and the intended recipient, if other than himself, of the nature of the material and the expected date and place of arrival.
2.
Prior to receiving radioactive materials from offsite, Radiation < hemistry shall verify that the quantity of radioactive material being received is such that the total quantity in the licensee's possession does not exceed the limit specified in his license for the particular isotope (s).
3 3.
Radiation-Chemistry must be provided with a copy of the radioactive materials shipment papers 'to determine the radiological precautions and special opening _ instructions if necessary before any package containing j
i radioactive material'is opened.
I 4.
Records of surveys performed on packages containing radioactive materials and on the transporting vehicle and radioactive material shipment papers i
shall be maintained, as required by 10 CFR 20, 70, and 71.
NOTE: Refer to 10 CFR 70 for the requirements for receiving special nuclear material.
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l February, 1982 COLLECTION, TPANSTER AND HANDLING OF RADIOACTIVE MATERIALS IN THE RADIOCHEMICAL LABORATORY AND COUNTING ROOMS Purpose This stancard provides guidelines for the collection, transfer and handling of radioactive materials in the Radiochemical Laboratory and Counting Rooms.
i General 1.
Before commencing work in these areas, personnel should be thoroughly trained in laboratory and counting room radiation safety measures.
This training should include instruction in the location and operation of all installed safety devices.
1 2.
All procedures must be approved as. required by the facility technical specifications before the procedure may be instituted.
3.
Contact Radiation-Chemistry for any additional protective equipment or dosimetry requirements (finger ring s, etc.) which may be applicable to a specific procedure.
- 4. _ Rules of general applicability for workers collecting, transferring, or handling radioactive materials in the radiochemical laboratory and counting rooms are as follows:
Containers used to transfer radioactive materials to the radiochemical a.
laboratory or to the counting room must be approved by the Radiation-Chemistry Supervisor.
Radioactive liquids should be moved in double non-breakable containers, b.
Protective equipment should be worn in plant sampling rooms and the radiochemical laboratory as required by Radiation-Chemistry.
c.
Protective equipment worn in ' the lab or counting room should be' monitored routinely for contamination.
d.
Personnel monitoring shall be performed as necessary so that working dose equivalent rates are known at all times, e.
Unless surveyed and released by Radiation-Chemistry, all equipment used in the radiochemical laboratory should be presumed to be contaminated.
page 57 of 69 i
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. February, 198 2 COLLECTION, TRANSFER AND HANDLING OF RADIOACTIVE MATERIALS IN THE RADIOCHEMICAL LABORATORY AND COUNTING ROOMS (CONTINUED) f.
Avoid direct csntact with radioactive materials by the use.of
- tweezers, tong s, or similar devices whepaver.
practical.
Mouth-operated equipment must-not be used in any area where radioactive materials are present.
g.
Appropriate shielding shall t,e used when handling or ~ storing radioactive materials where significant dose equivalent rates are involved.
Material stored for periods longer than 30 days shall be marked with a disposal cate.
h.
There is to be no smoking, eating, drinking, or chewing in the radiochemical laboratories or counting rooms.
5.
Any work area in which the possibility of high dose equivalent rates to workers having access to the area' exists shall be conspicuously posted if it becomes necessary to leave a setup unattended.
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February, 1962 RADIOACTIVE SOURCES Purpose This standard provides guidance relating to the proper use and storage of radioactive check and calibration sources.
Note:
The terms " check" And
" calibration" refer to the use of the source not to any characteristic or to j
the quantity of the source.
General 1.
The handling and use of instJ ument calibration sources by personnel other than Radiation-Chemistry personnel must be authorized by the Radiation-Chemistry Supervisor or Lead Health Physicist for each use.
2.
Any trained radiation teorker may routinely use instrument check sources as required.
I 3.
Instrument and area monitor calibrations must be conducted in accordance with written and approved procedures.
4.
Contact Radiation-Chemistry immediately if a source is lost' or otherwise i
removed from the direc.t control of the person using the source.
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5.
Each non-exempt check and calibration sou.rce must be leak tested at intervals not to exceed six months or as required by State and/or Federal regulations.
6.
All licensed radioactive sources shall be cataloged and identified by q
storage location or department using the source.
]
Portable Calibration Source Storage Containers 1.
Locked storage containers must be supplied for all portable calibration sources containing quantities of radioisotopes in excess of one millicurie.
i 2.
Portable containers must be constructed in such a manner as to reduce the radiation level to less than 5 mrem /hr at twelve inches f rom the external surface of the container.
Page 59 of 69
February, 1982 RADIOACTIVE SOURCES (CONTINUED) 3.
All storage containers for calibration sources shall be labeled in accordance with the regulations in 10 CFR Part 20.
In addition, labeling shall incluce information regarding isotope identification, date, ano quantity anc any additional information which may be helpful to the user such as: " Beta Source," " Neutron Source," etc.
CalibrationJacility 1.
Calibration sources are not to be removed f rom the calibration facility wells or from any calibration device or from the calibration facility area except upon approval of the Radiation-Chemistry Supervisor.
2.
The calibration facility must be kept locked when unoccupied.
Storage All licensed calibration sources will be stored in a locked room or a locked -
container.
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Page 60 of 69
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f February, 1982 MOVEMENT OF RADIOACTIVE MATERIALS IN THE STORAGE AND TRANSFER POOLS Purpose This standard provices spe:ific guidance pertaining to the movement of' radioactive materials in the storage and transfer pools.
Notes Contaminated materials should be handled as described in the Sectiot.,
" Storage of Contaminated Materials".
General-1.
Contact Radiation-Chemistry for protective equipment and monitoring requirements, if any, prior to the start of the job.
2.
An appropriate calibrated instrument must be employed when moving materials in the storage and transfer pools.
Restrictions 1.
Restrictions relating to the movement of radioactive materials in the pools ares a.
Materials giving a dose equivalent rate, including background, greater than 25 rR/hr at 6 inches from the surface of the object must be returned to the pool, and Raciation-Chemistry contacted for monitoring.
b.
Materials may not be lifted to.a level in or above the pools such that the dose equivalent rate, including background, exceeds 25 rR/hr at the nearest point of personnel occupancy without Radiation-Chemistry monitorirg.
2.
Restrictions relating to materials removed f rom the pools are:
a.
Materials must be contained to the extent that there is no reasonable possibility of spreading contamination to uncontrolled areas.
b.
Materials must be labeled as radioactive materiala (per 10 CFR 20) before being transferred to another controlled area.
c.
Materials shall be decontaminated or contained to prevent unnecessary contamination of the pool area.
d.
Contact Radiation-Chemistry for an evaluation when it is necessary: to accumulate contaminated materials removed from the pools.
l Page 61 of 69
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February, 1982 LAUNDRY / DRY CIIANING FACILITY i
Purpose This standard describes Radiation-Chemistry requirements for the operation of the station laundry / dry cleaning f acility.
1 l
General 1.
Contaminated garments should be segregated and laundered in batches of similat activity levels to avoid increasing the contamination of relatively " cleaner" protective clothing during the laundry cycle.
2.
Minimize handling of contaminated clothing to prevent or control I
protective clothing contamination and contamination of the laundry / dry cleaning facility.
3.
Use the installed ventilation contamination control equipment whenever practical.
4.
If the protective clothing cannot be decontaminated to acceptable levels, as determined by Radiation-Chemistry, the clothing shall be discarded as contaminated waste.
Requirements 1.
Protective equipment requirements for work in the laundry / dry cler-ing f
.\\
facility:
a.
Protective equipment requirements for operation of the laundry / dry cleaning facility shall be obtained from Radiation-Chemistry.
b.
Protective equipment requirements for f acility maintenance work will be determined by Radiation-Chemistry at the start of each job.
l Respiratory protection, as recommended by Radiation-Chemistry, must be c.
worn when emptying laundry bag s and when sorting contaminated I
protective clothing.
2.
Radiation-Chemistry monitoring requirements when working in the laundry / dry cleaning f acility:
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i Page 62 of 69 i
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-February, 1982'.
I.AUNDRY/ DRY CLEANING' FACILITY j
(CONTINUED)
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a.
Contact Radiation-Chemistry for monitoring whenever a dose equivalent rate greater than 25 mR/hr at six inches is detected.
b.
Contact Radietion-Chemistry whenever there is a suspected equipment' malfunction or whenever "drif t" in the survey instrument: response is suspected.
3.
Inspection and release requirements for la'undered/ dry cleaned protective
- clothing, s.
Laundered / dry cleaned protective clothing must. be inspected-for serviceability prior to release from the laundry / dry cleaning facility, b.
Protective clothing (non-folded) must' read less than 1 mR/hr above background at one inch as measured with a calibrated monitor, c.
Inner surf aces of waterproof boots ' and shoe rubbers must read less than 5 mR/hr above background at 2 inches as measured with a calibra'ted monitor.
4.
Inspection and release requirements. for laundered / dry cleaned respiratory equipment.
a.
Laundered / dry cleaned respiratory equipment must be inspected for I
serviceability prior to reissuance, b.
Fixed contamination on respiratory equipment must read less than 1 mR/hr above background at one inch.
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l Fage 63 of 69 l
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l February, 1982 SANDBLASTING CAVE
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This standard defines the Radiation-Chemistry requirements for
/
performing work in the sandblasting cave.
General 1.
Sandblasting, welding,
flame
- cutting, he ating,
sawing,
and' grinding i
operations on radioactive materials will be permitted within approved l
sandblasting caves without a Radiation Work Permit provided j
a.
Radiation-Chemistry has evaluated the job and established dose rates l
and protective clothing requirements.
t b.
All personnel within the cave wear supplied air respiratory equipment whenever the above operations are in progress.
When these operations are not in progress, filter-type respiratory protective equipment may I
be required for other work activities in the cave.
c.
The cave ventilation system has been dete rmined to be properly operating prior to the start of the job, d.
The worker's daily whole body dose equivalent does not exceed 50 mrem.
e.
The cave is decontaminated after each use to reouce spreading-radioactive materials outside of the cave.
2.
If it is suspected that the radiological conditions in the cave may change l
to such an extent as to lead to abnormal radiological conditions, leave the area immediately and notify supervision and Radiation-Chemistry.
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i Page 64 of 69
m___m_m_______
February, 1982 PERSONNEL RADIATION PROTECTION TRAINING Purpose This standard provides guidelines for radiation protection training of personnel performing work at CECO. nuclear f acilities.
General 1.
The extent of worker training shall be commensurate with the indivioual's duties and the radiological conditions in the work areas to which the individual may be assigned.
2.
Plant personnel shall be made aware of Commonwealth Edison's policy to maintain occupational radiation exposures as low as is reasonably achievable (ALARA) in accordance with 10 CFR 20.
Training 1.
All individuals whose duties entail working in or f requenting any portion of a restricted area or directing the activities of others who enter restricted areas shall be trained in accordance with the requirements set forth Th" 10 CFR 19.
Radiation protection training at the nucl'ar facilities shall also includes a.
Instruction in the use and limitations of radiation detectors with their direct application to self time keeping or dose assessment, b.
Instruction in the uses and limitations of respiratory and other protective devices employed at the facility; c.
Specific instruction concerning prenatal exposure risks to the developing embryo and fetus due to radiation or radioactive materials for female workers and those who may supervise or work with these individuals.
d.
A discussion of risk vs. the use of protective measures.
2.
All personnel who enter controlled areas of the station shall be retrained at intervals determined by the individual facility's Technical Specifications, not to exceed two years.
3.
Records of personnel ' training shall be maintained and updated as required by the Technical Specifications.
Page 65 of 69
February, 1982 REFERENCES The re f erence materials used throughout the standards are listed below.
These references were used ' to provide background information on the topics discussed in the Radiation Protection Standards.
With exception of regulatory requirements, inclusion of material on this list does not imply adherence to every rule stated in the individual reference.
1.
Title 10 CFR Part 19, " Notices, Instructions, and Reports to Workers:
Inspections."
2.
Title 10 CFR Part 20, " Standards for Protection Against Radiation."
3.
Title 10 CFR Part 30, " Rules of General Applicability to Licensing of.
Byproduct Material."
4.
Title 10 CFR Part 31, " General Licenses for Byproduct Material."
5.
Title 10 CFR Part 34,
" Licenses for Radiography and Radiation Safety Requirements for Radiographic Operations."
6.
Title 14 CFR Part 103 and Civil Aeronautics Board Publication Number 82,Section IV (1974);
U.
S.
Federal Aviation Regulations Concerning Radioactive Materials Transport.
i 7.
Titl'e 10 CFR Part 50, " Emergency Plans" i
i 8.
Title 10 CFR Part 70, "Special Nuclear Material."
9.
Title 10 CFR Part 71, " Packaging of Radioactive Material for Transport and Transportation of Radioactive Material Under Certain Conditions."
- 10. Title 39 CFR Part 124.25 and U.
S. Postal Service Publication Number 6 (1971);
U. S. Postal Regulations concerning transportation of radioactive materials.
- 11. Title 49 CFR Parts 170-189; cepartment of Transportation Regulations.
- 12. United States Nuclear Regulatory Commission (USNRC) Regulatory Guide 7.1,
" Administrative Guide for Packr.ging and Transporting Radioactive Materials."
Page 66 of 69
February, 1982 REFERENCES (CONTINUEDL
" Procedures for ~ Picking Up and Receiving Packages of Radioactive Material."
" Leakage Tests on Packages for Shipment of Radioactive Materials."
- 15. USNRC Regulatory Guide 7.4, " Administrative Guide for Obtaining Exemptions f rom Certain NRC Requirements Over Radioactive Material Shipments."
- 16. USNRC Regulatory Guide 8.3, " Film Badge Performance Criteria."
- 17. USNRC Regulatory Guide 8.4, " Direct Reading and Indirect Reading Pocket Dosimeters."
- 18. USNRC Regulatory Guide 8.8, "Information Relevant to Maintaining occupational Radiation Exposure as Low as is Reasonably Achievable."
" Acceptable Philosophy for Maintaining Occupational Radiation Exposures as Low as is Reasonably Achievable."
" Acceptable Concepts, Models, Equations, and Assumptions for a Bioassay Program."
- 21. USNRC Regulatory Guide 8.13, " Instructions Concerning Prenatal Radiation l
Exposure."
" Acceptable Program for Respiratory Protection."
- 23. USNRC NUREG-0041,
" Manual of Re spiratory Protection Against Airborne Radioactive Materials."
l
- 24. USNRC WASH-1313, " Procedures for Picking Up and Receiving Packages of Radioactive Material (1974)."
2$. State of Illinois Rules and Regulations for Protection Against Radiation.
- 26. American National Standard (ANSI) N2.1 - 1969, " Radiation Symbol."
- 27. ANSI N13.2-1969,
" Guide to Sampling Radioactive Materials in Nuclear racilities."
Page 67 of 69
February, 1982 REFERENCES (CONTINUE Q
- 28. ANSI N13.2-1969,
" Guide for Administrative Practices in Radiation Monitor ing. "
- 29. ANSI N13.5-1972,
" Performance Specifications for Direct Reading and Indirect Reading Pocket Dosimeters for X-and Gamma Radiation."
- 30. ANSI N13.6-1972, " Practice for Occupational Radiation Exposure Recoras Systems."
- 31. NASI N13.7-1972, " Film Badge Performance Criteria."
- 32. ANSI N14.5-1974, "LeakSge Tests on Packages for Shipment of Radioactive Materials."
- 33. ANSI N14.7-1975, " Guide to Design and Use of Shipping Packages for Type A Quantities of Radioactive Materials."
- 34. ANSI N14.10 1-1973, " Administrative Guide for Packaging and Transporting Radioactive Materials Shiprnents."
I
- 35. ANSI N14.10.2-1973,
" Administrative Guide for obtaining Department of Transportation Special Permits for Radioactive Materials fihipments."
- 36. ANSI N14.10-1975, " Administrative Guide for Ve rif ying. Compliance with Packaging Requirements for Shipments of Radioactive Materials."
- 37. ANSI N3U3-1978, " Internal Dosimetry for Mixed Fission and Activation Products."
l
- 38. ANSI Z88.2-1980, " Practices for Respiratory Protection."
- 39. National Council on Radiation Protection and Measurements (NCRP) Report Number 8 (National Bureau of Standards Handbook 48), " Control anc Regioval j
of Radioactive Contamination in Laboratories."
- 40. NCRP Report Number 30 (National Bureau of Standards Handbook 92), " Safe Handling of Radioactive Materials."
- 41. NCRP Report Number 39, " Basic Radiation Protection Criteria."
l
- 42. International.itomic Energy Agency (IAEA) Safety Standards, Safety Series Number 6,
" Regulations for the Safe Transport of Radioactive Material,"
l 1973 Revised Edition.
l Page 68 of 69
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l September 1985 REFERENCES (CONTINUED) 43.
IAEA Safety Standards, Safety Series Number 1, " Safe Handling of RaCAonuclides - 1973 Edition.
1 IFE Safety Standards, safety Series Number 38, " Radiation Protection j
44.
Ptocedures (1973)."
Occupational Safety and Health Administr ation (OSHA), Safety and Health 45.
Standards (29 CFR 1910), Section 1910.134, " Respiratory Protection."
46.
Generating Stations Emergency Plan (GSEP) and Emergency Plan Implementing Procedures.
1 47.
USNRC, I.E. Circular No. 81-07: Control of Radioactively Contaminated Material.
)
48.
USNRC, Richard C. DeYoung letter to E. Douglas Swartz dated May 18, 1982 concerning clarification of 1.E. Circular No. 81-07.
1 4
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- 2064H/5
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i ATTACHMENT F
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I.
9 AUXILIARY SYSTEMS Chapter 9, " Auxiliary Systems," for the Byron Station, Units 1 and 2, has been reviewed in accordance with the July 1981 edition of the " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants," (SRP),
.l NUREG-0800.
The staff has reviewed the design of the auxiliary systems necessary for safe reactor operation, shutdown, and fuel storage, or whose failure might-affeet plant safety, including their safety-related objectives and the manner in which these objectives are achieved.
~
l The auxiliary systems necessary for safe reactor. operation or shutdown include the essential service water system (ESWS), component cooling water system (CCWS), ultimate heat sink (UHS), control room chilled water system, the heating, ventilation, and air conditioning (HVAC) system for the control room and essential portions of the auxiliary building, essential portions of the chemical and volume control system (CVCS), and the auxiliary feedwater system (AfWS).
The auxiliary systems necessary to ensure the safety of the fuel storage f:cility include new fuel storage, spent fuel storage, the spent fuel pooi (acling and cleanup system, fuel handling systems, and the spent fuel pool arra i
ventilation system.
The staff has also reviewed other auxiliary systems to verify that their failure will not prevent safe shutdown of the plant or result in unacceptable release of radioactivity to the environment.
These systems include the nones.sen-tial service water system, the demineralized water makeup system, potable and j
sanitary water system, station heating system, nonessential cnilled water systems, nonessential portions of the CAS, nonessential portions of the CVCS, and HVAC system for nonessential portions of the auxiliary building, and the turbine building.
{
9.1 Fuel Storage and Handling 9.1.1 New-fuel Storage A single new-fuel storage facility located in the fuel-handling building it,
]
used for both Units 1 and 2.
The new-fuel storage f acility provides dry storage for 132 fuel assemblies (approximately 2/3 of a core load) and includes j
the new-fuel assembly storage racks and th! concrete storage vault that contains 1
the storage racks.
The fuel-handling building which houses the facility is designed to seismic Category I criteria, as are the storage racks and vault.
This building.is also designed against flooding and tornado missiles (refer to Sections 3.4.1 and
'3 3.5.2 of this SER).
Thus, the requirements of GDC 2 and the guidelines of Regulatory Guide 1.29 are satisfied.
Byron SER 9-1
J The new-fuel storage vault is not located in the vicinity of any high-energy lines or rotating machinery.
Physical protection by means of separation is provided for new fuel from internally generated missiles and the effects of-pipe breaks.
Therefore, the requirements of GDC 4 are met, as described in g
-]
Sections 3.5.1.1 and 3.6.1 of this SER.
Although the new fuel storage facility is common for both units, this sharing j
will not impair safety as discussed below and meets the requirements of G0C 5.
)
The new-fuel storage facility is designed to store unirradiated, low emission, fuel assemblies.
Accidental damage to the fuel would release relatively minor amounts' of radioactivity that'would be accommodated by the fuel handling.
~
The f acility is ' accessible to plant personnel for building ventilation system.
inspection.
Thus, the requirements of GDC 61, Fuel Storage and Handling and Radioactivity Control," are satisfied.
The new-fuel storage racks are designed to store the fuel assemblies in an array' with a minimum center-to-center spacing of 21 in., which is sufficient to maintain a K of less than 0.95 with the fuel assemblies flooded with~ cold, of 0.98 or less unboratedwaf$f TheracksaredesignedtomaintainaK'[8verisprovidedover r.
under optimum moderation (foam, mist, or fog).
A locked each new-fuel storage cell as additional protection for the new fuel. -The-racks themselves are designed to preclude the inadvertent placement of'a fuel assembly in other than the prescribed spacing.
Thus, the r. requirements of GDC 62, " Prevention of Criticality in Fuel Storage and Handling,"'are satisfied.
Based on its review, the staff concludes that the new fuel storage facility is in conformance with the requirements of GDC 2, 5, 61, and 62 as they relate to protection against natural phenomena, shared systems, radiation protection,:and prevention of criticality and the guidelines of Regulatory Guide 1.29 relating to seismic classification.
It is, therefore, acceptable.
9.1.2 Spent-Fuel Storage A single spent-fuel storage facility located in the fuel-handling building is shared by both plant units.
The spent-fuel storage f acility provides under-water storage f or 1050 f uel assemblies plus 10 failed fuel assemblies, or
{
approximately five full-core loads.
The facility includes the spent fuel storage racks and the lined spent-fuel storage pool that contains the storage racks.
The structure housing the facility (the fuel-handling building) is designed to seismic Category I criteria, as are the stainless steel storage racks and storage pool (including the gates between the pool), fuel transfer canal, and i
The stainless steel spent-fuel pool liner plate is designed cask loading area, to stay in place in an SSE, thus preventing potential mechanical damage to the spent fuel or damage resulting f rom overheating as a result of blocking of j
The fuel-handling building is also designed against i
cooling water flow paths.
flooding and tornado missiles (refer to Sections 3.4.1 and 3.5.2 of this SER).
Thus, the requirements of GDC 2 and the guidelines of Regulatory Guides 1.13, 1.29, and 1.117 are satisfied.
1 l
The fuel pool is not located in the vicinity of any high-energy lines'or rotating machinery.
Therefore, protection of speat fuel from internally 9-2 Byron SER
-.._.__._._._._m.______---__,
4
generated missiles and the effects of pipe breaks is previded by physical separation (refer to Sections 3.5.1.1 and 3.6.1 of this SER).
Thus, the
(
requirements of GDC 4 and the guidelines of Regulatory Guide 1.13 concerning missile protection for spent fuel are satisfied.
Although the spent-fuel storage f acility is common for both units, this sharing is acceptable as only one unit is refueled at a time.
It thus meets the requirements of GDC 5.
The seismic Category I storage rack arrangement provides a fuel storage array below 0.95 for both adequatetomaintainthemultiplicationfactor,K'N,afuelassembly.
normal storage and in case of accidental dropping The design of the storage racks is such as to preclude placement of a fuel assembly in a position other than the prescribed location.
The racks can withstand the impact of a dropped fuel assembly without unacceptable damage to the fuel and can withstand the maximum uplift forces exerted by the fuel-handling machine.
Thus, the requirements of GDC 61 and 62 and the guidelines of Regulatory Guide 1.13 concerning the protection of fuel from mechaaical damage and pre-vention of criticality are satisfied.
The design of the storage pool includes a pool water level monitoring system and radiation and temperature monitoring systems with local. indication and alarm in the control room.
These features satisfy the requirements of GDC 63.
I The staff has reviewed the compatibility and chemical stability of the mate-rials (except the fuel assemblies) wetted by the pool water.
The pool liner, rack lattice structure, and fuel-storage tubes ara stainless steel, which is compatible with the storage pool environrr.ent.
In this environ-ment of oxygen-saturated borated water, the corrcsive deterioration of the type 304 stainless steel should not exceed a depth of 6.00 x 10 5 in. in 100 l
years (Johnson,1977), which is negligible relative to the initial thickness.
I Dissimilar metal contact corrosion (galvanic attack) between the stainless steel of the pool liner, rack lattice structure, f uel-storage tubes, and the Inconel and the Zircaloy in the spent-fuel assemblies will not be significant because all of these materials are protected by highly passivating oxide films and are, therefore, at similar potentials.
From the evaluation as discussed above, the staff concludes that the corrosion that will occur in the spent-fuel storage pool environment should be of little significance during the 40 year life of the plant.
Components in the spent-fuel storage pool are constructed of alloys that have a high resistance to general corrosion, localized corrosion, and galvanic corrosion.
The staff, therefore, concludes that the environmental compatibility and stability of the materals used in the spent-fuel storage pool are adequateThe based on test data and actual service experience in operating reactors.
staff finds that the selection of appropriate materials of construction by the applicant meets the requirements of GDC 51, because there is the capability to permit appropriate periodic inspection end testing of components, and GDC 62, because there is the capability to prevent criticality by maintaining structural integrity of components and is, therefcre, acceptable.
1 I
Byron SER 9-3 l
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9 4
Based on its review, the staff t.oncludes that the spent-fuel storage facility is in'conformance with the requirements of GDC 2, 4, 5, 61, 62, and 63 as they relate to protection against natural phenomena, missiles, pipe break effects, shared systems, radiation protection, prevention of criticality, and monitoring provisions, as well as the guidelines of Regulatory Guides-1.13, 1.29, and 1.117 concerning the facility's design, seismic classification,.and protection again.ct tornado missiles, 2nd is, therefore, acceptable.
9.1.3 Sper.t-Fuel-Pool Cooling and Cleanup System The spent-fuel pool cooling and cleanup system is shared by both units.
It is designed to remove the decay heat generated by the stored spent-fuel assemblies and to maintain clarity and purity of the spent-fuel pool water._ The spent-fuel pool cooling system is designed to remove the decay heat generated by the number of spent-fuel assemblies that are stored following a back-to-back refueling of both station units and the accumulated fuel assemblies from pre-vious annual cyclic refuelings.
The system includes all components and piping from inlet to exit from the storage pools, piping used for fuel pool makeup, I
and the cleanup filter /denineralizers to the point of discharge to the radwaste system.
The design consists of two essential fuel pool cooling trains, each f
with a spent fuel pit pump and heat exchanger that are completely redundant.
A separate nonsafety-relsted fuel pit skimmer pump and filter is also provided f or keeping the pool water surf ace clean.
The essential portions of the system are housed in the seismic Category I, flood-ano tornado protected fuel-handling building (refer to Sections 3.4.1 and 3.5.2 of this SER).
The system itself, with the exception of the cleanup portion, is designed to Quality Group C and seismic Category I requirements.
failure of the nonsafety-related (nonseismic Category I) Quality Gro m 0 cleanup portion will not affect operation of the cooling train as isolation capability of that portion of the piping system is provided, and, therefore, no adverse effect on safety-related equipment would result from such a failure.
Therefore, the design satitifies the requirements of GDC 2 and the guidelines of Regulatory Guides 1.13, 1.26, ano 1.29 with respect to seismic and quality aroup classification of the fuel pool cooling system.
Discussion of compliance with Regulatory Guide 1.52, " Design, Testing, and Maintenance Criteria for Engineered-Safety-Feature Atmosphere Cleanup System Air filtration ano Adsorp-tion Units of of Light-hater-Cooled Nuclear Power Plants," for the fuel-handling building is discussed in Section 9.4.2 of this SER.
The various components of the system are located in separate missile-shielded cubicles within the tornado-missile protected fuel-handling building and are separated from other moderate-and high energy piping systems (refer to Sec-tions 3.5.1.1 and 3.6.1 of this. SER).
Thus, the requirements of GDC 4 are satisfied.
Although the spent-fuel pool tooling and cleanup system is common for both units, this sharing will not impair safety as a single failure in the system does not adversely effect the system safety function for either unit.
- Thus, the requirements of GDC 5 are satisfied.
Either fuel pool cooling tra5n maintains the pool water temperature at 126 F or less with a heat load based on decay heat generation for the design storage Byron SER 9-4
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ATTACHMENT G
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b/e-I S AE AMENDMENT 38' 1
MAY 1982 Ihe sp+nt tuel bridge.crar.e transters r.ew Iuel f rom the new f uel
]
elevator to the upending cevice end spent Iuel f rom 'the apendir.g device into a spent f uel storage rack.
Ine ne w f uel is transf erreo Irom tne new f uel rachs to the r.ew-
~
f uel elevator cy the f uel handling ou11 ding crane.-
Ihe rus1 har.oling area planoviea is 'anown in Figure 1. 2-9 : ar.o in
-)
sectional elevatien in F igure 1.2-11.
l In spent f uel handling operation is carried out entirely under water.
'h Ine nes ruel nandling operation ist alsc normally-carried out i
i anaer sater except during the ir.terval when the new f uel is
']
transIerreo f rom the snippirig contaiaers to the new tuel storece ossioly during initial core area or r.ew t uel elevator, or e ioaalng.
j l
Initial core fuel with up to 3.20'w/o U-235 enrichment may be
.)
stored dry in the spent fuel storage racks in a checkerboard pattern only, prior to'the initial core loading, utilizing administrative controls to prevent improper loading.. Detailed j
discussions are given in Subsection 9.1.2.
~he Iuel handling Luilding whien encloses the sper.t ruel pool, new r uel storage area and eortions or the tuel trandrer' system is a seismic Categor y I tuilding.
It is therefore capatle of withstanding tne design-cas ts ear t hquakes, tornados, ar.d missilas, and is not sucject to floo3ing.
Cetails of the seismic d esigr. ar e given in Sucse ction 3. o. 4.1. 2.
9.1.1
!4w Fuel Storace 9.1.1.1 Design Bas es New fuel is otorea ary in racks aesig:ied to provide stordge f or 13 2 t u+1 assembiles.
The cedign of tne normally cry new storage r o ci.o is suen that the ef fective maltiplicatior, f actor does r.ot exceed the design-casis limit 'or 0.9e witn I ael of a maximum-enrichment of 4.00 w/o U-23 5 in pls es, assuming optimum neutror.
l moae ra tion conditior.s (dry st fogged) in accoraance with AN81 N1s.2-1973.
The nes ruel stor age racxs ar e desagt.ed to precl ude. storage et a r uel asamLly other than where int endea, ar.a to withstand tr.e upiatt Ivrce wnich could occur due to ruel assemtly hanging up daring litting and tne impa ct lead of a aroppeo' f uel' assembly.
Ine ne. tuel storage f acility is designed to prevent floodir.g.
N e ve r tt.e l e s s, tne new f uel storage racas are oeoigr.ed such that tna e f f ec ti ve multiplication. f actor dass not exceed 0.95 with tuel or a n.aximum enrichment of
- 4. 00 w/o U-235' in p lace, assuming the stored assemtl es complete 2y cabmergea ir. ur.boratec water at a conservative water temperature una with r.o credit icr neutror.
ol out. i r. te.e I ut i assembly.
r 9.1-2 L- -
B/B-FSAR AMENDMENT 38 MAY 1982 The new fuel storage racks are designed to withstand the safe shutdown earthquake and design-basis wind and. tornado loadings.
For further information see Subsections 3.2.1, 3.3.1, and 3.3.2.
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9.1-2a o
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i B/B-FSAR AMENDMENT 43
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SEPTEMBER 1983 9.1.1.2 Facility Description New fuel assemblies and control rods are received in an area of the fuel handling building separate f rom other fuel and are stored in racks in the new fuel storage vault af ter removal from the shipping containers.
The fuel for the initial core loading is temporarily stored in the spent i
fuel pit.
This area is. shown in Figures 1.2-9 and 1.2-11.
-l New f uel assemblies are stored in racks in a reinforced concrete pit located in the fuel handling building as shown in Figure 1.2-9.
The new fuel storage pit provides dry storage for 132 fuel assemblies which is approximately 2/3 of a core.
The new fuel storage installation is shown in Figure 9.1-1.
Each rack cc~sists of individual lockable
'j i
vertical cells of stainle-
.meel construction.
A group of 44 cells, in rows of t-
,t a center-to-center spacing
{
of 21 inches is assembled tv che support structure.
The support structure is bolted and/or welded to embedded studs in the concrete walls and columns of the storage pit.
The j
fuel assembly rests on a self-leveling support plate.
All l
i surfaces that contact fuel assemblies are austenitic stain-less steel.
The supporting structure is high strength carbon l
steel.
The racks are designed to withstand normal operating loads as well as design-basis seismic loads.
The ability to handle off-normal refueling sequences is assured by the quantity of the spent fuel storage racks I
which will handle 1050 f uel assemblies.
The total' storage I
capacity is 1050 fuel assemblies in the spent fuel pool, f
10 f ailed fuel assemblies in the spent fuel pool, and 132 1
new fuel assemblies in the new fuel storage racks, i
The new fuel storage racks rest on an 18-inch floor slab with 2-foot thick reinforced concrete walls.
Access to the area is through openings in the floor at print elevation 426 feet 0 inch.
The new fuel storage racks are laterally supported by attachment to embedment plates in the walls.
9.1.1.3 Safety Evaluation 1
l Criticality Control l
The design of the new fuel storage racks provides for a suberitical effective multiplication factor (k for the optimum moderation condition of less than 0.98 kn)d for the f
flooded condition of less than 0.95 in accordance with ANSI i
N18.2-1973 and ANSI N210-1976, respectively.
The optimum moderation condition exists when the new fuel racks are l
covered with clean water at a uniform density of 0.018 gm/cc.
9.1-3
B/B-FSAR AMENDMENT 38 MAY 1982 i
The flooded condition exists when the new fuel racks are covered with cold, clean, unborated water with a full density.
l The criticality analyses of these two conditions included several conservative assumptions as well as the effects of uncertainties in the calculation method, and geometric and material variations of the fuel storage rack.
The follow-ing conservative assumptions were used in the calculation:
a.
Fresh fuel of 4.00 w/o U-235 enrichment - Initially i
the maximum enrichment will be lower than this, I
but it could approach this enrichment if an IB-month fuel cycle is used.
b.
The neutron leakages in the vertical direction l
and in one of the radial directions (X) are neglected while the neutron leakage in the other radial direction (Y) is accounted for.
c.
Fixed neutron absorbers in the fuel assembly and soluble poison in the water in case of flooding are neglected.
The calculations to determine the optimum moderation condition were performed with a Monte Carlo neutron transport code, KENO, with a 123-group cross section library generated from a basic GAM-THERMOS library using the NITAWL routine in j
the AMPX code package.
The effects of small changes in i
rack parameters on rack system criticality were determined with a four-group diffusion theory code, PDO-07, and with the neutron cross sections generated from the NUS Corporation version of LEOPARD, NUMICE.
Both PDO-07 and the KENO cal-culation methods have been benchmarked.
Under the optimum moderation condition of 0.018 gm/cc water with full storage racks, the k as determined from KENO is 0.9250.
Calculational unce?b$inties were determined from both comparison between calculation and experiments using KENO, and a statistical evaluation of Monte Carlo runs.
The results are a calculation uncertainty for the former of 0.0086 ok and, for the latter, 0.0089 Ak, at a 95% confidence level; or a total calculation uncertainty of 0.0175 ok.
Variation in material composition and mechanical spacing and tolerances acting in a direction to increase the system ofgb.were also analyzed and resulted in a reactivity increase k
0342 ok.
Adding reactivity effects of calculation uncertain-ties of 0.0175 ok and geometric and material uncertainties of 0.0342 ok to the optimum moderation k,gg results in a maximum 9.1-3a
t
' AMENDMENT'38 B/B-FSAR MAY 1982 k
of 0.977 with a 95% probability at a 95% confidence 188$1, which satisfies the design basis of maintaining k,gg below the design-basis limit of 0.98.
When the full new fuel storage racks'are. flooded with cold
.is 0.8655.
Adding reactivity unborated water, the nominal k'bdies of 0.0173 ak, geometric effects of calculation uncertaT
. and material uncertainties of-0.0134 ok, and the temperature increase.effect (from 68' F to 170' F) of 0.0024 ok to the results in a maximum k of 0.899 with a 95%
nominal k probabiliIffata95%confidencelevI$fwhichsatisfiesthe design basis of maintaining k,gg below the design-basis limit of 0.95.
9.1-3b.
A m________m___m._._._.-_-___..__...
_. _ _. ~ _
I B/B-FSAR AMENDMENT 19 MARCH 1979 New Fuel Storage Fack Structural Design In accordance with requirements for seismic design classifica-tion, the new fuel storage racks are classified Seismic Category I.
Structural integrity of the fuel racks when suojected to design-oasis seismic loads has been demonstrated.
The loads, load combinations, ano structural acceptance criteria (strength limits) that were considered are identical to those l
considered for the spent and failed f uel storage racks, (Subsection. 9.1. 2), except that thermal loads were considered negligicle and therefore not included.
l Ihe seismic loading of the new fuel rack is determined from.a I
response spectrum modal dynamic analysis in which the stif fness of tne f uel assemoly is neglected.
However, the mass of the fuel assemolies is considered to be uniformly distributed along the storage tubes in the north-south and east-west directions.
The entire mass of the fuel is assumed to De supporteo by the floor in the vertical direction.
Racks are modeled in detail using finite elements.
Since the cans and fuel rest on the floor, a two-dimensional model representing either the upper or lower support system is used for the vertical earthquake.
-Ihe can is vertically rigid and was analyzed for vertical earthquakes.
The seismic analysis is performed using the STARDYNE computer program.
Io determine the earthquake response, SI AEDYNE is rirst run to determine the natural f requencies and participation I actors.
For frequencies with significant modal participation, mode shapes and modal loads are calculated using the appropriate response spectra.
Closely sl aced modes are comcined directly and then combined in a square root of the sum of the squares (SRSS) manner with other significant modes.
The results of the three directions of earthquake are combined in an SRSS f ashion as directed in Combining Modal Responses and Spatial Components in Seismic Response Analysis (Regulatory Guide 1.92).
Using the previously listed loads and load combinations, stresses were calculated at critical sections of the racks.
The results of the structural and seismic analyses demonstrate that the fuel racks are structurally adequate and meet the design criteria.
Since rack integrity is maintained there would be no damage to the stored fuel assemblies and no increase in Eeff uncer these loads.
New Fuel Rack Desian Features l
l The new fuel racks were designed and f abricated with a high degree of reliability and integrity.
The codes and standards used for the new fuel racks are the same as those used for the spent and l
failed fuel storage racks and are listed in Subsection 9.1.2.
9.1-4
'b/B-F3Ah AMENDMENT 38-MAY.1982 9.1.2 frent Fuel Storage i
9.1.1.1 Desigr. Baoes a:1 stat aae space is proviceo in.the spent luel storage pool for 1050. fuel assemblies and 10 damaged _ fuel-assemblies.
.ne sper,t f uel racx desagr. plecluaes stolage of a ruel assembly a t' locatior.s in the lacA other that enose designeo.to' receive R
a dae molle s.
Cha center-tv-center spacir.g tet ween'~ stored f uel assemclies in a l
I ally loaded spent. ruel stcrage. rsco 16 suf ficient to malr.t ain. a.
and
'ef f esa 1 to or less than 0. 95 tori tne normal wet lconcition, l
Ior all abnormal and accider.t coniitions.
Ihis' design: Lasis is
-l met with fresh fuel or up to an enrichnent of. 4. 00 - w/o ' U-2 3 5, a conservative water temperature (ebo F), no credit for fixed neutron polsan in the tuel assemLly or soluole neutron poison in tne pool sater, and no credit f or structural material other than the stainless steel cans.
Tne s er.t fuel rocks and tr.e f ailed f uel canisters were destgr.(3 c
to allow aceq uate cooling or tna s :nt tuel assemtlies.
Both tr.e e
l spant ano failed fuel racxs are classirled seismic category 1.
l
.ney are designed to withstar.o the ef tects or the' Sbi, remain f ur.ctional and maintain suLcriticality.
Ine spent f uel ra cks were des agt.ea to withstar.d either a dropp( d f uel assembly or the upwaro turce of - 3 stucx assemtly without loss or f unction.
The struct ural analysis is discussed later J
l in this section.
Snieloing Ior the s ent fuet stcrage estrangement is s af tidier.t to e
rvtect elant personnel trom exposure to as low as reasonably eacnievacle and well below 10 Cfh 20 limits.
- ne spent f uel storage f acility was designea to prevent torr.aco-generated missiles from causir.g damage to tne f uel.
)
9.1.2.e Facilities Description l
Scent F u e l st or a g e F a ct s j
Ine sper.t tu;l storage racxs irovide a place in the spent tuel starage pool f or dtvr ing tr.e sper.t rael uischarged. Iron,the reactor vess el.
They are top (ntry racts, designeo to mair. tait.
the sper.t Iuel ir. a space gecaetry tnat precluaes the possibility or criticality under t.oth normal ana str.ormal conditior.s.
l The oesign or the spent ruel storage racx assemL1) preve nts ar.y l
F ossioility of accidental criticality.
ne location or the sper.t f ael pool withir. tne plant is snown in r igure 1.1-9.
A gener al
.i a rra r.g emen t of spent ruel storage tac 111 ties is shovr. Ar. liour+
9.1-2.
)
i v.1-5
3 1
B/r-F3Ah AMEMDMENT 3B MAY 1982 2he sient Iuel storage racxs (snown in 11gurc 9.1-3) consist of square etainless steel tutes or 1/6 iach thicx 30u austenitic stainless steel.
They are held at 14 inenes cer.ter-to-cer.ter r y tye 304 stainless steel plateo.
Ihe plates, which are also 1/ e inc ti.ick, are welded to the sices of the sguare storage tutes ot four elevations.
Ine tunes are Ilsted at tha top to permit f
easy storage ar.d retrieval of the stored fuel assemL11es, ano to be comt atible witn the f uel handling equipment.
Two different rack arrays are utilized to maximize ase or the availatie tuel ace in trae pool.
i ne racx structure is welded to on storage or elovat ed case which is a system or selded teams ano stifrenars.
Trr: La de se rves to support the weight or the rual assemolic s an.d to alstricate the loao on tr.e pool 11oor.
The Lase contains an opening at each fuel assembly storage location wriich accommodates tne Iael assemoly lower r.ozzle.
.Jatural circulation of pool water down between the storage tuoes 2na up through the lower r.oz.le or eacn ruel assemtly removes aecay neat.
The storage esils ar e designeo to provice lateral support for storec assamolles 01 the Westingbouce 17 x 17 array design, ano other a s era.11e s witu the same external aimer.sions ano similar lower nozzle design.
To prev'ent raovement of the racks, particularly from seismic ex-citation, the racks are bolted to embedment plates in the floor of the spent fuel pool.
The racks are raised off the floor on adjustable legs to allow natural circulation flow.
The legs are adjusted after installation to 1cvel the racks and distribute the load to all legs. Refer to Figures 9.1-20 and 9.1-21 for embeiment nlate details and plan, j
i The spent fuel storage racks can also be used for dry storage of I'
initial core fuel with up to 3.20 w/o U-235 enrichment in a checkerboard pattern only, prior to the initial core loading.
Thc safety evaluation for this mode of fuel storage is given in Subsection 9.1.2.3.
e li _Iuei_atorace Packs d
t o-1 assena lies that ar e dama ge d or Ia11+d such tnat they may re At as-radioactivity ar* stor ed in special containers in tne t alled f uel r acx.
ine toiled t u+1 stolage racx is locatec 1r. the tuel pool ar.o p.ovices stvinge Ior 10 t oilec Iuel o p sr.t assemtlles.
Th16 racx is s !.own it. Figurc 9.1-4.
Its Cor.51sts of 1/u inen u.ack stair.2 t ss steel tuu-s spaced at 22 inches center-to-cenier, and support d by stainless steel plat es along their 1+ngtna.
ihe r acx etruct ur t is welaei to an elevated case to sapport tne w:1ght ci t ne r at l ass + ro lieo ur. tne pcol 11001.
i t. e Lase cor.tains an oge:. Ar.g at each 8tursge locatior. tc allow natural circ ala tior, of t ool wat er ae tr. rough this operir.q to remove ascay h'at trom tne stcred assently.
To pre ver.t nov:mer.t o r ts.e ra ce 6,'
a c s 1..
particularly y due to melsmic e xcitation, this racx is alos ;olt ed to eca.edme r.t platts
- 11. ti.e Iloor or the s; nt t u( 1 pool.
T1.e rack also has oni raise the lacxs ott the ad3 2st apl e le gs to level ine raci.o rio;r to =110w not ar al circulation 14aw.
9.1-o
l B/b-FSAR AMENDMENT.43 SEPTEMBER 1983 6 pent Fuel Storage Pool
)
1 Ihe spent f uel storage pool is designed' f or the underwater storage of spent tuel assemolies and control rods af ter their removal from the reactor.
The pool depth is such that the surf ace dose level will De approximately 2.5 mrem /hr when moving a fuel assemoly over storage racks.
It is designed to accommodate a total of 1050 ruel assemolies.
Tne spent fuel pool f
general arrangements are shown in Figures 1.2-9 and 1.2-11.
j f uel assemolies are handled oy a fuel handling tool Seent and suspended from an overhead monorail electric hoist manipulated by an operator standing on a. movable bridge over the pool.
The spent fuel storage pool and transrer canal are constructed of structural reinforced concrete and lined with stainless steel.
A new ruel elevator is located in the. pool to allow transferring new f uel assemblies into tne pool tor suosequent handling with j
the spent f uel pool handling crane.
The elevator is equipped witn alarms to inrorm the operator of malf unction of the elevator during movement of new fuel.
The elevator carriage is always.
celow tne level of water in the pool.
All parts in contact with tne pool water are stainless steel.
This elevator is a
3 II-structurally designed as Seismic Category A se pa ra te, walled-oft area is proviaed at the end of the pool for storing spent fuel cssks.
The cask storage area is provided with leaktight gates to allow isolation f rom the pool.
The design of the fuel handling ouilding crane rails prec1udes trovel of the crane's hook over the spent fuel storaca racks except for storage of fuel assemblies for initial core lording.
For operations following initial core loa 6ing, the only lifting device which can travel over the spent. fuel racks is the s9ent fuel pit bridge crane hoist.
This hoist is equipped with a load-limiting interlock to prevent liftino a load her. vier than 4000 pounds.
If the load is in excess of this weig.t, cne electrical interlock will stop the fuel bridge winch drive from moving upward.
Fuel assemblies received for initial core loading nay be inserted directly into the spent fuel storage racks 'using the new fuel handling tool and the fuel handling building crane.
This method of fuel storage will require an override of the fuel handling building crane interlocks and would occur only for receipt and dry storage of first core fuel.
Ine design requirements for the separating wall between the fuel cas4 pit and the spent fuel pool are:
Walls are designed to withstand increased water a.
4 pressure wnich may De caused oy a vertical drop of I
the cask.
The wall thickness is 5 feet.
9.1-7
B/B-FSAR AMENDMEWT'43 SEPTEMBER 1983 For a cask. drop on the exterior pool wall, the wingi ts.
.iall will he 'alloLed to - def orm locally ceyond = its elastic' limit, cut it nas oeen shown.that the cask' will.not. fall outside the. cask storage:well"and.thus will not. af f ect the' f uel in the. f uel storage pool.
4 i
i 4
1 l
i f
i
. i 9.1-7a i
1
B/B-FSAR AMENDMENT 46 JANUARY 1985 c.
Since the cask will. always. be. brought to the pool so as to' enter the storage well, without passing over the pool, there is no possibility of cask-drop into the fuel storage pool.
d.
The spent fuel pool floor is 6-foot thick reinforced concrete resting directly ~ on bedrock (Byron) or on soil (Braidwood) and is designed for a vertical drop of the cask in the storage well.
Gates are provided to transfer f uel from the pool to e.
the spent fuel cask pit and to the f uel transfer canal.
9.1.2.3 Safety Evaluation Criticality Control The design of the spent fuel storage racks provides for a suocritical ef fective multiplication f actor (keff) for both normal and abnormal storage conditions less than 0.95 in accordance with ANSI N210-1976.
Normal conditions exist when the fuel racks are covered with a normal depth of water (at least 23 feet above the top of stored fuel) for radiation shielding, and with the maximum number of fuel assemblies ~in their design storage position.
An abnormal condition may result from accidental dropping of a fuel assemoly or horizontal movement of a fuel assembly to a location adjacent to a loaded storage rack.
The criticality analyses of the normal condition included several conservative assumptions as well as the ef f ects of uncertainties in the calculation method, and geometrle and material variations of the fuel storage rack.
The following conservative' assumptions-were used in the calculation:
Fresh fuel of 4.00 w/o U-235 enrichment - Initially a.
the maximum enrichment will be lower than this, but.
it could approach this enrichment if an 18-month fuel cycle is used.
The enrichment selected is higher than the average enrichment of any fuel expected to be stored in the spent fuel pool for conservatism.
b.
Spent f uel pool bulk water temperature - A calculation was done to determine the increase in reactivity due to a f urther decrease in the design-basis temperature of 680 F and this reactivity was added to the nominal keff for still further conservatism.
c.
The neutron leakages in all three directions were neglected.
d.
Fixed neutron absorbers in the fuel assembly and soluble poison in the pool water are neglected.
9.1-8
3 AMENDMENT 38 B/B-FSAR MAY 1982-The majority of the calculations were performed with methods commonly used in light water reactor design, i.e.,
four group diffusion theory with all calculations using PDO-07.
Cross sections for these. calculations are generated with NUMICE, the NUS Corporation.versi'on.of the Westinghouse LEOPARD code.
This code uses the same cross section library tape and calculation techniques as LEOPARD.
Selected cases were checked and the final design-multiplication factors were verified with Monte Carlo calculations using KENO with a 123-group cross section library generated from a basic GAM-THERMOS library using the NITAWL routine in the - AMPX ~
l code package.
Both the PDQ-07 and the KENO calculation methods as described above have been benchmarked.
the k Under normal conditions with full storage racks, CalculationuncertIkkties l
as determined from KENO is 0.9085.
were determined from both comparison between calculation and experiments using KENO, and a statistical evaluation of Monte Carlo runs.
The results are a calculation uncertainty for the former of 0.0086 ok and, for the latter, 0.0087 ok, l
at a 95% confidence level; or a total calculation uncertainty of 0.0173 ok.
Variation in material composition and mechanical j
spacing and tolerances acting in a direction to close the water gaps between adjacent racks were also analyzed and resulted in a reactivity increase of 0.0134 ok.
As previously l
stated, the effect of a temperature decrease was also included.
A temperature increase to 170* F resulted in a reactivity increase of 0.0024 ak.
Adding reactivity effects of calculation uncertainties of 0.0173 ok, geometric and material uncertainties of 0.0134 Ak, and the temperature decrease effect of 0.0024 tk to the nominal k
, results in a maximum k of 0.942 with a 95% proba6[$ity at a 95% confidence T$hel, which satisfies the design basis of maintaining k,ff'below the design-basis limit of 0.95.
Two abnormal conditions were also considered.
They are (1) a dropped fuel assembly assumed to lay across the top of fuel rack and (2) a fuel assembly in transport in a vertical position accidentally dropped into a position parallel with the stored fuel in the most reactive corner among the racks.
Of these two, the second is most severe, as the end fittings on top of the fuel assembly prohibits the dropped fuel. assembly on top of the rack from being nearly as close to the stored fuel as the assembly beside the rack.
The analytical techniques and assumptions used for the analyses of the second case are the same as used for the analysis of the normal condition except that credit was taken for a minimum 1950 ppm soluble boron to bring the initial conditions into the range of 4
credibility.
The analysis is still conservative as the i
conservatism for water temperature reduction was not removed, 1
even though credible water temperature is well above 32' F.
I 9.1-9 j
AMENDMENT 38 B/B-FSAR MAY 1982 i
Thus, the I
worst accident case as analyIbb,well below 0.762.
The results show a nominal k does not have consequence any more than the worst normal distribution case.
I Dry Storage of New Fuel in Spent Tuel Racks The design of the spent' fuel storage racks also allows a suberitical effective multiplication factor for dry storage of new fuel with 3.20 w/o enrichment in a checkerboard loading pattern: k less than 0.98 for the optimum moderation conditionIkkkin accordance wTkb.less than 0.95 for the flo ANSI.N210-1976.
Under the checkerboard loading pattern, new fuel assemblies.are to be placed in rack positions with four nearest adjacent positions empty.
The optimum moderation condition exists when the storage racks are covered with clean' water at a uniform' density of 0.05 gm/cc.
The' flooded condition exists when the storage.
racks are covered with cold, clean, unborated water at a full density.
The criticality analyses of these twc conditions were performed with the assumptions and calculational methods similar to those utilized for the analyses of new fuel racks in Subsection 9.1.1.3.
The only exception was that the spent fuel storage rack was assumed to be infinite in size in all three dimensions for additional conservatism.
Under the optimum moderation condition of 0.05 gm ec water for the spent fuel racks containing the 3.20 w/o enrichment fuel in the checkerboard loading-pattern, the k as deter-mined is 0.8904.
Addingreactivity.effectsofbkkculational uncertainties of 0.0153 ok and geometric and material uncer-results taintiesof0.0479Aktotheoptimummoderationk,kka95%
in a maximum k of 0.954 with a 95% probability confidencelev$$fwhichsatisfiesthedesignbasisofmain-taining k,gg below the design-basis limit of 0.98.
When the. spent fuel racks containing the 3.20 w/o enrichment fuel in the checkerboard loading pattern are flooded with the nominal k is 0.8413.
Adding cold, unborated water, reactivityeffectsofcalculationun8kftaintiesof0.0198ok, 8
geometric and material uncertainties of 0:0056 ok, and the temperature decrease effect (from 68* F to 40' F) of 0.0014 ok to the nominal k results in a maximum k of0.868witha95%probabTk(tyata95%confidenceTkhel, which satisfies the design basis of maintaining k,f f -below the design-basis limit of 0.95.
9.1-9a 1
_ _ _ _ = _ _ - _ _ _ _ _ _ -
B/B-FSAR AMENDMENT 19 MARCH 1979 Fuel Storage Rack Structural Design In accordance with the spent fuel storage facility design Lases and the requirements for seismic design classification, the spent and f ailed f uel storage racks are designed as Seismic Category I.
Structural integrity of the racks when subjected to normal, abnormal, and seismic loads was demonstrated.
consequently, the following loads, loao combinations, and structural acceptance criteria are considered:
Dead loa'ds - the load due to dead weight of tne rack, a.
f uel assemblies, and ouoyancy.
L.
Live loads - the load due to lif ting empty racks during installation.
c.
Thermal loacs - tne load due to un1Iorm thermal expansion of the racxs caased by a cnange in average pool temperature Ircm 400 I to 1500 F, a thermal gradient between adjacent storage locations of 180 F l
for spent fuel storage racks, and 220 I for failed l
fuel storage racks.
d.
Severe environmental load - the load due to the operating basis earthquake (oBE).
e.
Extreme environmental load - the load due to the sate shutdown earthquake (SSE).
1.
Accidental drop of a tael assembly f rom the maximum possible height consistent with fuel handling operations.
9 a postulated stucx fuel assemoly causing an upwaro 1
force equal to tne ruel grapple load limit exerted on tne assembly during attempted withdrawal.
Ine f uel storage racks were analyzed asing the elastic worxing stre ss aesign methods for tne Iollow ing load combinations:
a.
deaa loads plus live load, l
o.
cead loads plus OBE loaos, c.
cead loads elus normal tnermal loads plus obE loads, d.
dead loads plus normal thermal loads plus SSE 1 cads, e.
dead loads plus nornal tnermal loads plus f uel assembly drop, and l
l 9.1-10 w _- -
-B/B-FSAR AMENDMENT 32 AUGUST 1981 f.
dead loads plus normal thermal loads'plus stuck fuel assembly load.
Live loads are not included in load combinations b through f, since the only live load on the rack occurs during lifting.
Lifting of the racks.is performed only when the racks are empty.
The following strength limits apply for the above load combinations:
Load Combination Strength Limit 1.0S a
b 1.0S 1.5S c
d 1.65 1.6S (except as noted e
below) f 1.6S - (except as noted below) l l
Where S is the required section strength based on the j
)
elastic design methods and the allowable stresses defined l
J in Part 1 of the AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings,"
February 12, 1969, including Supplement Numbers 1, 2, and 3.
(Supolement 3 was ef f ective, June 12, 1974, and revised June 30, 1975.)
For load combinations e and f, local stress-es might exceed the limits, orovided there is no loss of function of the fuel rack.
J Both the spent fuel and f ailed fuel storage racks were an s-lyzed to determine that these strength limits are not exceeded, a.
Spent Fuel Storage Racks The seismic loading of a fuel rack. module is l
determined from a response spectrum-modal dynamic
?
analysis in which the stiffness of the fuel as-sembly is neglected.
However, the mass of the fuel assemblies and an effective mass of water are considered to be uniformly distributed along the storage tubes, The appropriate floor response spectra and damping values (2%_and 4%)'for the l
OBE and DBE, respectively, are employed.
The STARDYNE computer program is c used to perform the structural analysis of the racks.
Racks are modeled in detail' using beam and - plate finite 9.1-11
l B/B-FSAR AMENDMENT 19' MARCH 1979 1
i I
elements.
The three-dimensional. finite element model i
I for a' spent fuel rack is shown in Figure:9.1-5.
1 To determine the earthquake response,1 STARDYNE is' first run to determine the natural frequencies-and participation factors.. For frequencies,with significant modal participation, mode shapes, and l
1 1
i i
J
,1 l
l J
i I
]
I a
.i 1
\\
l l
l I
i 9.1-11a i
4
1 B/B-FSAR modal loads are calculated.
Closely spaced modes are combined directly and then combined in an SRSS manner with other significant modes.
The results of the three directions of earthquake are combined in an SRSS f ashion per requirements f or combining modal responses and spatial components in seismic response analysis.
In tne general seismic / structural analysis of the fuel racxs, the mass of a f uel assembly is assumed to be uniformly distributed along the length of each of the fuel storage cans.
This assumption is conservative because the lower racx fundamental frequencies are calculated using a relatively stiff rack design, resulting in a higher seismic amplified acceleration loading on the rack.
Since a gap of approximately 1/4 inch exists between the sides of a fuel assembly and the can, the fuel will actually move within the can during a seismic event and cause impact loads to be transmitted to the fuel rack.
The eff ects of this fuel-can interaction were analyzed by utilizing the ANSYS computer l
program.
A nonlinear dynamic analysis of a single can and fuel assembly was performea to determine the shear force and bending moment which may occur at
}
critical sections of the can as a result of the fuel assembly impacting the can at the maximum velocity.
Tne can and fuel assembly were modeled by finite elements separated cy nonlinear gap elements as shown j
in Figure 9.1-6.
The can has stiffness characteristics representative of a can within a rack.
The fuel, whien was assumed to be pinned at its base (by friction), was given an initial velocity relative to the can.
This initial velocity is equal to the SRSS summation of the floor velocity and the velocity of the rack with respect to the floor.
D.
Failed Fuel Racks The seismic loading of the f ailed f uel rack was determined from a response spectrum modal dynamic analysis in which the stif fness of the fuel assemoly and failed fuel container was neglected.
- Again, however, the mass of the f ailed f uel container and its contents and an ef fective mass of. water are considered to Le uniformly distributed along the storage tunes.
The appropriate response spectra for the OBE and SSE were employed.
The STARDYNE computer program was used to perform the structural analysis of the racks.
Failed fuel racks were modeled in detail using beam and plate elements.
The three t
9.1-12
B/B-FSAR RMENDMENT 19 MARCH 1979 dimensional finite element model for the failed fuel rack is shown in Figure 9.1-7.
As for the spent tuel storage racks, to determine the earthquake response, STARJYNE was first run to determine the natural fre u ?ncies and participation.
v factors.
For frequencies with significant modal participation, mode shapes and modal loads were calculated.
Closely spaced modes were comhined directly and then combined in an SRSS manner with other significant modes.
Ihe results of the three directions of earthquake was combined in an SRSS fashion per requirements for combining modal responses and spatial components in seismic response analysis.
In the seismic and structural analysis of the fuel racks, the mass of a failed fuel container and its contents was assumea to oe uniIormly distributed along the length or each of the f uel storage cans.
Ihis assumption is conservative because the lower rack fundamental f requencies calculated loading on the racx due to the relatively stiff rack design, result in higher seismic amplified acceleration.
A gap on the order of 1/4-inen exists cetween the sides of a failed fuel container ano the can, thus the container will actually move within the can during a seismic event and cause impact loads to ce transmitted to the racx.
The effects of this fuel-can interaction were considered cy conservatively doacling the results of the f requency response analysis to account ror tne effects of impact.
Using the previously listed loads and load combinations, stresses are calculated at critical sections of the racks.
The results of the structural and seismic analyses demonstrate that the fuel racks are structurally adequate and meet the design criteria.
Since rack integrity is maintained, no damage to stored assemblies and no increase in K w uld occur under these loads.
eff Fuel Rack Design Features Both the spent and failed fuel storage racks are all stainless steel, as is the spent fuel pool liner, to minimize the potential for galvanic corrosion.
Stainless steel has also been shown to be compatible with spent fuel pool water and the stored assemblies.
The fuel rack base is elevated above the floor to assure adequate i
flow under the rack in each fuel assembly.
Analyses have been performed and show that sufficient flow is induced by natural convection to preclude local boiling in the hottest storage location.
The analyses were based on the following assumptions.
{
4 9.1-13
]
B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 1
a.-
The fuel element inlet-tempe rature is ' the mixed hot design temperature of. the pool.
b.
A hot assembly peaking f actor ot - 1.55 is applied to the core average assembly. energy release rate of 2.3 x 105 Btu /hr.
The maximum local peaking tactor is 2. 33, ' giving a c.
maximum local neat Ilux or 1706 Btu (hr-f t2).
6.
A film coefficient or 39 Btu (hr-ft2-0 F) is cased on pure conduction. through a stagnant boundary layer at the fuel rod surface.
A downcomer region between adjacent cans-within the e.
rack completely feeds two assemolies, each assumed to be generating the maximum' heat rate defined in I
assumption b.
f.
One dimensional fluid flow analysis applies.
During Iull core of fload with the culk pool temperature at a design value of 1500 F, the mixed temperature of the water exiting rrom tne hottest storage location is less than 1860 F.
l Ihis is 540 F Delow the local saturation _ temperature of 2400 F inalcating tnat adequate margin to culk bolling. exists.
Under design operating conditions, the f ael rod surf ace temperature calculated on tne casis of the heat flux and film coefficient defined aoove, is 110 F below the local saturation temperature and thus precludes local coiling.
Other spent Fuel Storage Facility Design Features l
O The Byron /Braidwood stainless steel spent fuel pool _ liners were originally designed and constructed to Category I requirements.
Weld inspections conducted subsequent _to the initial acceptance of the Byron liner did not confirm the acceptability of-those welds to Category I standards.
The liners at both plants were.
therefore reclassified in April 1983 to Category II structures.
Because they meet the additional criteria of NUREG-0800, fuel Revision 3, July.19 81, the liners for the rpent f uel pool, transfer canal, and spent fuel cask pit do not need to be designed j
and erected to Category I requirements.
Long term benefits in
]
terms of reduced record keeping and surveillance requirements 1
also result from this action without degrading safety.
l The design of the liner is such that the design. thermal stresses
~
are much larger than the design seismic (SSE) stresses.
Because of this fact, and because the liner was designed and constructed 1
J as a Category I structure, gross failure of the liner as a result of a seismic event is not considered to be credible.
Thus, neither mechanical damage to fuel nor flow blockage of spent fuel in the spent fuel racks as a result of liner failure are considered to be credible events, l
9.1-14
]
j
i AMENDMENT 43 B/'B-FSAR SEPTEMBER 1983 1
r A failure of the liner which allows leakage of the pool (such as the rupture of a seam) as a water past.the liner result of a seismic event would not have any adverse effects in on fuel in the pool or on any safety-related equipment the plant.
Five 1 inch drains are provided behind the spent fuel pool 11iner.
The drain piping is embedded in the concrete structure up-to column-row W, where it joins the auxiliary building-floor. drain system.
At the. point-where the drain piping emerges from the concrete wall, five normally closed (one per drain line) are provided.
Any valves and sight glasses the liner would be collected by the drain piping and leakage past No other leakage paths exist.
The auxiliary stopped by the valves.
building floor drain piping is seismically. supported, so failure of this piping as a result of an SSE is not postulated..Thus, failure of the liner would not result in loss of significant quantities of pool water, and no safety-related equipment would be affected by any resulting leakage.
The activity release trom a dropped or ruptured fuel assembly l
shall be limited such that the radiation dose at the site coundary will not exceed the limits stated in 10 CFE 100.
fuel Activity releases in the event of fuel damage in the spent pool would be limited by the decontamination f actor provided by tne pool water.
The ventilation system provided in the f uel handling building is equipped with filtration devices to limit the potential release of radioactive materials including iodine.
The design is descrioed in detail in Subsection 9.4.2.
Ine f uel storage f acility is contained and the equipment is designed so that accidental releases of radioactivity are monitored and will not exceed the guidelines of 10 CFE 100.
Subsection 15.7.4 contains analyses of hypothetical. fuel handling accidents.
The fuel rac4s are designed and fabricated with a high degree of reliability and integrity.
A list of codes and standards used r or tne spent and f ailed fuel storage racks is given celow:
9.1-14a
_g o
a B/B-FSAR 1
a.
Design Codes 1.
AISC Manual of~ Steel Construction, 7th Edition, 1970; and ATSC " Specification for the Desigr, Fabrication 2.
and Erection of Structural Steel f or! Buildings,"
February _ 12, 1969'and Supplements 1, 2, and 3.
(Supplement 3 effective June 12, 1974).
3.
ANSI N18. 2-1973, Nuclear' Saf ety Criteria for.the Design of Stationary Pressurized Water Reactor Plants.
4.
ANSI N210-1976, LRR' Spent Fuel Facilities.
b.
M_ aterial Codes 1.
ASME Specification.SA-240, Specification for Stainless and Eeat Resisting Chromium and Chromium-Nickel Steel Plate Sheet and Strip for Fusion-Welded Unfired Pressure Vessels; 2.
ASME Specification SA-230 Specification for Alloy Steel Bolting Materials for Low Temperature Service; and 3.
ASME Specification SFA-5.9, Corrosion Resisting-k Chromium and Chromium Nickel Steel-Welding Rods
}
and Base Electrodes.
f c.
Welding Codes 1.
ASME Boiler and Pressure Vessel Coce Section j
IX-1974, Wslaing and Erazing Qualifications.
d.
Quality Assurance, Cleanliness, and Packaging j
?
Requirements Commonwealth Edison Company and/or other approved l
1.
]
[
quality assurance requirements; 1
2.
ANSI N45.2.2, Packaging, Shipping, Feceiving, Storage and Bandling of Items for Nuclear Power Plants; ana 1
3.
ANSI N45. 2.11, Quality Assurance Requirements for l
the Design of Nuclear Power Plants.
9.1.3 Spent Fuel Pool Coolina and Cleanup System Ihe spent fuel pool cooling system (SFPCS) is designed to remove fuel assemolies from the i
the decay heat generated cy stored spent spent f uel pool water.
This cooling is accomplished by taking
)
9.1-15
,