MNS-18-045, Emergency Plan Implementing Procedure HP/0/B/1009/002, Revision 003, Alternative Method for Determining Dose Rate within the Reactor Building
| ML18207A422 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 07/24/2018 |
| From: | Duke Energy Carolinas |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18207A387 | List: |
| References | |
| MNS-18-045 HP/0/B/1009/002, Rev. 003 | |
| Download: ML18207A422 (31) | |
Text
EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0)
AD-EP-ALL-0602 Rev.4 ATTACHMENT 4 Page 1 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form>>
Screening and Evaluation Number Applicable Sites BNP D
EREG #: 2193370 CNS D
CR3 D*
HNP D
MNS X
5AD #: 2193356 ONS D
RNP D
GO D
Document and Revision HP/O/B/1009/002 Alternative Method for Determining Dose Rate within the Reactor Building Rev 003 Part I. Description of Activity Being Reviewed (event or action, or series of actions that may result in a change to the emergency plan or affect the implementation of the emergency plan):
AR02080957, the primary change to this procedure is to change the location of the survey used to determine dose rates within the reactor building when 1 or 2 EMFs 51A/B are inoperable during emergency conditions. The survey location is being changed from on contact with the outside surface of the outer upper airlock to on contact with the outside surface of the Spent Fuel Building to Upper Containment VE Door. With the additional distance from the containment source and additional shielding provided by the VE door the multiplication factor used in Step 4.1.3 and Enclosure 5.1 is changed from 740.0 to 1363. The variables used in determining the new multiplication factor are included with the procedure as Enclosure 5.2. Location of survey is being changed because it is unlikely that a survey of the outer airlock door will be possible with VE in operation. The fuel building location is simpler, safer and doable with VE in service.
\\
Instrument names Teletector and Extender have been deleted. Teletector and Extender are trade names which are no longer used here.
References to survey instrument MCHPS number have been deleted and simply replaced with instrument number.
Additisinal changes: Inserted Revision History page, restored procedure format Part II. Activity Previously Reviewed?
I Yes ID I No I X
EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0)
AD-EP-ALL-0602 Rev.4 ATIACHMENT4 Page 2 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form>>
Is this activity Fully bounded by an NRC approved 10 CFR 50.90 submittal or 10 CFR 50.54(q)
Continue to Alert and Notification System Design Report?
Effectiveness,
Evaluation is not 10 CFR If yes, identify bounding source document number or approval reference and required. Enter 50.54(q}
ensure the basis for concluding the source document fully bounds the proposed justification Screening change is documented below:
below and Evaluation complete Form, Part Ill Justification:,
Part V.
Bounding document attached (optional}
D Part Ill. Editorial Change Yes lo No X
Is this activity an editorial or typographical change only, such as formatting, 10 CFR 50.54(q)
Continue to paragraph numbering, spelling, or punctuation that does not change intent?
Effectiveness,
Evaluation is not Part IV and Justification:
required. Enter address non justification and editorial complete changes,
Part V.
Part IV. Emergency Planning Element and Function Screen (Reference Attachment 1, Considerations for Addressing Screening Criteria)
Does this activity involve any of the following, including program elements from NUREG-0654/FEMA REP-1 Section II? If answer is yes, then check box.
1 10 CFR 50.47(b}(1) Assignment of Responsibility (Organization Control) 1a Responsibility for emergency response is assigned.
D 1b The response organization has the staff to respond and to augment staff on a continuing basis D
(24-7 staffing) in accordance with the emergency plan.
2 10 CFR 50.4 7(b )(2) Onsite Emergency Organization 2a Process ensures that onshift emergency response responsibilities are staffed and assigned D
2b The process for timely augmentation of onshift staff is established and maintained.
D 3
10 CFR 50.47(b)(3) Emergency Response Support and Resources 3a Arrangements for requesting and using off site assistance have been made.
D 3b State and local staff can be accommodated at the EOF in accordance with the emergency plan.
D (NA for CR3) 4 10 CFR 50.47(b)(4) Emergency Classification System 4a A standard scheme of emergency classification and action levels is in use.
D (Requires final approval of Screen and Evaluation by EP CFAM.)
5 10 CFR 50.47(b)(5) Notification Methods and Procedures
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(J EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
AD-EP-ALL-0602 Rev.4 Sa Sb Sc ATTACHMENT 4 Page 3 of 5
<< 10 CFR 50.54(q) Screening Evaluation Form>>
Procedures for notification of State and local governmental agencies are capable of alerting them of the D
declared emergency within 15 minutes (60 minutes for CR3) after declaration of an emergency and providing follow-up notification.
Administrative and physical means have been established for alerting and providing prompt instructions D
to the public within the plume exposure pathway. (NA for CR3)
The public ANS meets the design requirements of FEMA-REP-10, Guide for Evaluation of Alert and D
Notification Systems for Nuclear Power Plants, or complies with the licensee's FEMA-approved ANS design report and supporting FEMA approval letter. (NA for CR3)
EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0)
AD-EP-ALL-0602 Rev. 4
<< 10 CFR 50.54(q) Screening Evaluation Form>>
ATIACHMENT 4 Page 4 of 5 Part IV. Emergency Planning Element and Function Screen (cont.)
6 10 CFR 50.4 7(b )(6) Emergency Communications 6a Systems are established for prompt communication among principal emergency response D
organizations.
6b Systems are established for prompt communication to emergency response personnel.
D 7
10 CFR 50.47(b)(7) Public Education and Information 7a Emergency preparedness information is made available to the public on a periodic basis within the D
plume exposure pathway emergency planning zone (EPZ). (NA for CR3) 7b Coordinated dissemination of public information during emergencies is established.
D 8
10 CFR 50.47(b)(8) Emergency Facilities and Equipment Ba Adequate facilities are maintained to support emergency response.
D 8b Adequate equipment is maintained to support emergency response.
D 9
10 CFR 50.47(b)(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use.
X 10 10 CFR 50.47(b)(10) Protective Response 1 Oa A range of public PARs is available for implementation during emergencies. (NA for CR3)
D 1 Ob Evacuation time estimates for the population located in the plume exposure pathway EPZ are available D
to support the formulation of PARs and have been provided to State and local governmental authorities. (NA for CR3) 1 Oc A range of protective actions is available for plant emergency workers during emergencies, including D
those for hostile action events.
1 Od Kl is available for implementation as a protective action recommendation in those jurisdictions that D
chose to provide Kl to the public.
11 10 CFR 50.47(b)(11) Radiological Exposure Control 11a The resources for controlling radiological exposures for emergency workers are established.
D 12 10 CFR 50.47(b)(12) Medical and Public Health Support 12a Arrangements are made for medical services for contaminated, injured individuals.
D 13 10 CFR 50.47(b)(13) Recovery Planning and Post-accident Operations 13a Plans for recovery and reentry are developed.
D 14 10 CFR 50.47(b)(14) Drills and Exercises 14a A drill and exercise program (including radiological, medical, health physics and other program areas)
D is established.
14b Drills, exercises, and training evolutions that provide performance opportunities to develop, maintain, D
and demonstrate key skills are assessed via a formal critique process in order to identify weaknesses.
14c Identified weaknesses are corrected.
D 15 10 CFR 50.47(b)(15) Emergency Response Training 15a Training is provided to emergency responders.
D
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EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
<< 10 CFR 50.54(q) Screening Evaluation Form>>
Part IV. Emergency Planning Element and Function Screen (cont.)
16 10 CFR 50.47(b)(16) Emergency Plan Maintenance 16a Responsibility for emergency plan development and review is established.
AD-EP-ALL-0602 Rev.4 AITACHMENT 4 Page 5 of 5 D
16b Planners responsible for emergency plan development and maintenance are properly trained.
D PART IV. Conclusion If no Part IV criteria are checked, a 10 CFR 50.54(q) Effectiveness Evaluation is not required, then complete D, 10 CFR 50.54(q) Screening Evaluation Form, Part V.
Justification:
If any Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV criteria are checked, then complete X, 10 CFR 50.54(q) Screening Evaluation Form, Part V and perform a 10 CFR 50.54(q)
Effectiveness Evaluation. Program Element 4a requires final approval of Screen and Evaluation by EP CFAM.
Part V. Signatures:
EP CFAM Final Approval is required for changes affecting Program Element 4a. If CFAM approval is NOT required, then mark the EP CFAM signature block as not applicable (N/A to indicate that signature is not required.
Reviewer Na e (Print):
Renard 0. Bums Date:
"5/15J1s Approver (EP Manager Name (Print):
Kevin L Murray Approver (EP CFAM, as required) Name Date: ;/,~
QA RECORD
EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
AD-EP-ALL-0602 Rev. 4 ATTACHMENT 5 Page 1 of 7
<< 10 CFR 50.54(q) Effectiveness Evaluation Form>>
Screening and Evaluation Number Applicable Sites BNP D
EREG #:2193370 CNS D
CR3 D
HNP D
MNS X
SAD #:2193356 ONS D
RNP D
GO D
Document and Revision HP/O/B/1009/002 Alternative Method for Determining Dose Rate within the Reactor Building Rev 003 Part I. Description of Proposed Change:
AR02080957, the primary change to this procedure is to change the location of the survey used to determine dose rates within the reactor building when 1 or 2 EMFs SlA/B are inoperable during emergency conditions. The survey location is being changed from on contact with the outside surface of the outer upper airlock to on contact with the outside surface of the Spent Fuel Building to Upper Containment VE Door. With the additional distance from the containment source and additional shielding provided by the VE door the multiplication factor used in Step 4.1.3 and Enclosure 5.1 is changed from 740.0 to 1363. The variables used in determining
, the new multiplication factor are included with the procedure as Enclosure 5.2. Location of survey is being changed because it is unlikely that a survey of the outer airlock door will be possible with VE in operation. The fuel building location is simpler, safer and doable with VE in service.
Instrument names Teletector and Extender have been deleted. Teletector and Extender are trade names which are no longer used here.
References to survey instrument MCHPS number have been deleted and simply replaced with instrument number.
Additional changes: Inserted Revision History page, restored procedure format, 10 CFR 50.54(q) Initiating Condition (IC) and Emergency Action Level (EAL) and EAL Yes Bases Validation and Verification (V&V) Form, is attached (required for IC or EAL change)
No X
EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
<< 10 CFR 50.54(q) Effectiveness Evaluation Form>>
Part II. Description and Review of Licensing Basis Affected by the Proposed Change:
DESCRIPTION MNS Emergency Plan Section I ACCIDENT ASSESSMENT rev 18-1 1.6 Release Rates/Projected Dose For Offscale Instrumentation AD-EP-ALL-0602 Rev. 4 ATIACHMENT 5 Page 2 of 7 If instrumentation used for dose assessment is offscale or inoperable, dose rates within the Reactor Building will be determined using procedure HP/O/B/1009/002, Alternative Method for Determining Dose Rate Within the Reactor Building, or HP/O/B/1009/006, Procedure for Quantifying High Level Radioactivity Release During Accident Conditions.
The change to HP/O/B/1009/002 continues to meet the requirement of MNS Emergency Plan Section 1(6)
Release Rates/Projected Dose For Offscale Instrumentation Part Ill. Description of How the Proposed Change Complies with Regulation and Commitments.
If the emergency plan, modified as proposed, no longer complies with planning standards in 10 CFR 50.47(b) and the requirements in Appendix E to 10 CFR Part 50, then ensure the change is rejected, modified, or processed as an exemption request under 10 CFR 50.12, Specific Exemptions, rather than under 10 CFR 50.54(q):
NU REG 06541 (6). Each licensee shall establish the methodology for determining the release rate/projected doses if the instrumentation used for assessment are offscale or inoperable.
10CFRS0.47(b)(9) Adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use.
MNS Procedure HP/O/B/1009/002, Alternative Method for Determining Dose Rate Within the Reactor Building, describes and provides a method to determine dose rates in the Reactor Building during an emergency situation when lEMFSl A&B or 2EMF51 A&B are not available. Revision 2 of the procedure was approved on 11/23/1999 and required the user to obtain a dose rate at the 'Upper Personnel Hatch Air Lock Door (R/hr)' described in step 4.1.3. The current perspective is that it would be extremely difficult to open the VE (Annulus Ventilation) Door at the Upper Containment (U/C) access point and doing so would also present a significant personnel safety concern.
A preferential location would be the outside of the VE Door from the associated Spent Fuel Building near the U/C access. The travel path and this location are more easily accessible, exposes the performer to lower dose rates and does not subject the user to unnecessary safety risks. The selected location at the Spent Fueling Building side of the VE door is also directly aligned with the airlock doors to provide practical dose rate information. The new calculated correction factor was derived using the same methodology as the previous factor but is at a farther distance and accounts for additional shielding encountered by the additional VE door.
MNS continues to maintain methodology for determining the release rate/projected doses if the instrumentation
EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0)
<< 10 CFR 50.54(q) Effectiveness Evaluation Form>>
AD-EP-ALL-0602 Rev.4 ATIACHMENT5 Page 3 of 7 used for assessment are offscale or inoperable. The changes to HP/O/B/1009/002 provides a safer more efficient method to determine dose rates in the event (1)(2) EMF 51A & EMFSlB is offscale or inoperable.
These proposed changes continue to support the requirements described in 10CFRS0.47(b){9) and NUREG 06541 (6).
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EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0)
AD-EP-ALL-0602 Rev. 4 ATIACHMENTS Page 4 of 7
<< 10 CFR 50.54(q) Effectiveness Evaluation Form>>
Part IV. Description of Emergency Plan Planning Standards, Functions and Program Elements Affected by the Proposed Change (Address each function identified in Attachment 4, 10 CFR 50.54(q) Screening Evaluation Form, Part IV of associated Screen):
AD-EP-All-0602 Attachment 5 10 CFR 50.54(q) Effectiveness Evaluation Form states:
9 10 CFR S0.47(b)(9) Accident Assessment 9a Methods, systems, and equipment for assessment of radioactive releases are in use.
NUREG 0654 NUREG 0654 I. Accident Assessment Planning Standard states "Adequate methods, systems and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use."
NUREG 0654 I (6). Each licensee shall establish the methodology for determining the release rate/projected doses if the instrumentation used for assessment are offscale or inoperable.
10CFRS0.47{b}(9) Adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use.
MNS continues to maintain methodology for determining the release rate/projected doses if the instrumentation used for assessment are offscale or inoperable. The changes to HP/O/B/1009/002 provides a safer more efficient method to determine dose rates in the event (1)(2) EMF 51A & EMFSlB is offscale or inoperable as described in NUREG 0654 I (6)..
These proposed changes continue to support the requirements Adequate methods, systems and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use as described in 10CFR50.47(b)(9).
Part V. Description of Impact of the Proposed Change on the Effectiveness of Emergency Plan Functions:
Part V. Description of Impact of the Proposed Change on the Effectiveness of Emergency Plan Functions:
MNS Procedure HP/O/B/1009/002, Alternative Method for Determining Dose Rate Within the Reactor Building, describes and provides a method to determine dose rates in the Reactor Building during an emergency situation when 1EMF51 A&B or 2EMFS1 A&B are not available. Revision 2 of the procedure was approved on 11/23/1999
EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0)
AD-EP-ALL-0602 Rev. 4 ATIACHMENT 5 Page 5 of 7
<< 10 CFR 50.54(q) Effectiveness Evaluation Form>>
and required the user to obtain a dose rate at the 'Upper Personnel Hatch Air Lock Door (R/hr)' described in step 4.1.3. The current perspective is that it would be extremely difficult to open the VE (Annulus Ventilation) Door at the Upper Containment (U/C) access point and doing so would also present a significant personnel safety concern.
A preferential location would be the outside of the VE Door from the associated Spent Fuel Building near the U/C access. The travel path and this location are more easily accessible, exposes the performer to lower dose rates and does not subject the user to unnecessary safety risks. The selected location at the Spent Fueling Building side of the VE door is also directly aligned with the airlock doors to provide practical dose rate information. The new calculated correction factor was derived using the same methodology as the previous factor but is at a farther distance and accounts for additional shielding encountered by the additional VE door.
MNS continues to maintain methodology for determining the release rate/projected doses if the instrumentation used for assessment are offscale or inoperable. The changes to HP/O/B/1009/002 provides a safer more efficient method to determine dose rates in the event (1)(2) EMF 51A & EMFSlB is offscale or inoperable as described in NUREG 06541 (6)..
These proposed changes continue to support the requirements Adequate methods, systems and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use as described in 10CFR50.47(b)(9).
The change to HP/O/B/1009/002 continues to meet the requirement of MNS Emergency Plan Section 1(6)
Release Rates/Projected Dose For Offscale Instrumentation.
The proposed changes described in Revision 003 of HP/O/B/1009/002 Alternative Method for Determining Dose Rate within the Reactor Building do not result in a reduction in effectiveness of facilities, response organizations, or response equipment. The proposed changes do not reduce the effectiveness of The McGuire Emergency Plan or the Emergency Plan Implementing Procedure HP/O/B/1009/002.
Part VI. Evaluation Conclusion.
Answer the following questions about the proposed change.
1 Does the proposed change comply with 10 CFR 50.47(b) and 10 CFR 50 Appendix E?
YesX 2
3 4
a Does the proposed change maintain the effectiveness of the emergency plan (i.e., no reduction in effectiveness)?
Does the proposed change maintain the current Emergency Action Level (EAL) scheme?
Choose one of the following conclusions:
Yes X YesX The activity does continue to comply with the requirements of 10 CFR 50.47(b) and 10 CFR 50, Appendix E, and the activity does not constitute a reduction in effectiveness or change in the current Emergency Action Level (EAL) scheme. Therefore, the activity can be implemented without prior NRC approval.
No 0 No 0 No D X
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'v EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(Q)
<< 10 CFR 50.54(q) Effectiveness Evaluation Form>>
AD-E P-ALL-0602 Rev.4 ATTACHMENT 5 Page 6 of 7 b
The activity does not continue to comply with the requirements of 10 CFR 50.47(b) or 10 CFR 50 Appendix E or the activity does constitute a reduction in effectiveness or EAL scheme change.
D Therefore, the activity cannot be implemented without prior NRC approval.
Part VII. Disposition of Proposed Change Requiring Prior NRC Approval Will the proposed change determined to require prior NRC approval be either revised or I Yes D I No D rejected?
If No, then initiate a License Amendment Request in accordance 10 CFR 50.90 and AD-LS-ALL-0002, Regulatory Correspondence, and include the tracking number:
EMERGENCY PLAN CHANGE SCREENING AND EFFECTIVENESS EVALUATIONS 10 CFR 50.54(0)
AD-EP-ALL-0602 Rev. 4 ATTACHMENT 5 Page 7 of 7
<< 10 CFR 50.54(q) Effectiveness Evaluation Form >>
Part VIII. Signatures: EP CFAM Final Approval is required for changes affecting risk significant planning standard 10 CFR 50.47(b)(4). If CFAM approval is NOT required, then mark the CFAM signature block as not applicable (N/A) to indicate that signature is not required.
Reviewer Na e (Print):
Renard 0. Bu Date:
Approver (EP ~~- Nrut=¥ajPrint):
S-?-LJ-1 Date:
Approver (CF AM, as require ) Name (Print):
rV illt;l If the proposed activity is a change to the E-Plan or implementing procedures, then create two EREG General Assignments.
If required by Section 5.6, Submitting Reports of Changes to the NRC, then create two EREG General Assignments.
One for EP to provide the 10 CFR 50.54( q) summary of the analysis, or the completed 10 CFR 50.54(q),
to Licensing.
One for Licensing to submit the 10 CFR 50.54(q) information to the NRC within 30 days after the change is put in effect.
QA RECORD D
D
Duke Energy Company Procedure No.
MCGUIRE NUCLEAR STATION HP/O/B/1009/002 Alternative Method For Determining Dose Rate Within the Reactor Building Revision No.
003 Ref ere nee Use PERFORMANCE I
This Procedure was printed on 6/26/2018 9:02 AM from the electronic library as:
(ISSUED) - PDF Format Date(s) Perfonned I Work Order/Task Number (WO#)
COMPLETION I
d Yes D NA Checklists and/or blanks initialed, signed, dated, or filled in NA, as appropriate?
D Yes D NA Required attachments included?
D Yes D NA Charts, graphs, data sheets, etc. attached, dated, identified, and marked?
D Yes D NA Calibrated Test Equipment, if used, checked out/in and referenced to this procedure?
D Yes D NA Procedure requirements met?
Verified By Date
- Printed Name and Signature Procedure Completion Approved Date
- Printed Name and Signature Remarks (attach additional pages, if necessary)
IMPORTANT: Do NOT mark on barcodes.
Printed Date:
- 6/26/18*
Attachment Number:
- FULL*
I IIIIII IIII I II Ill lllll llll I II 1111111111111111 1111111111111111111 IIIII II IIII Revision No.:
- 003*
11111111111111111 IIII IIII Procedure No.:
- HP/O/B/1009/002*
I IIIIIII IIIIII II I II Ill llll I II IIII Ill I II Ill lllll lllll lllll llll I II Ill 111111111111111111
Duke Energy Procedure No.
McGuire Nuclear Station HP/0/B/1009/002 Revision No.
Alternative Method for Determining Dose Rate 003 within the Reactor Building EPIP procedure. Issuance of procedure must be coordinated with Emergency Planning to ensure resources are available to update hard copies of procedures.
Reference Use Electronic Reference No.
MC0045FV
Revision History (significant issues, limited to one page)
HP/0/B/1009/002 Page 2 of 4 Rev 003 (12/8/2016) AR02080957, the primary change to this procedure is to change the location of the survey used to determine dose rates within the reactor building when 1 or 2 EMFs 5 lA/B are inoperable during emergency conditions. The survey location is being changed from on contact with the outside surface of the outer upper airlock to on contact with the outside surface of the Spent Fuel Building to Upper Containment VE Door. With the additional distance from the containment source and additional shielding provided by the VE door the multiplication factor used in Step 4.1.3 and Enclosure 5.1 is changed from 740.0 to 1363. The variables used in determining the new multiplication factor are included with the procedure as Enclosure 5.2.
Instrument names Teletector and Extender have been deleted.
References to survey instrument MCHPS number have been deleted and simply replaced with instrument number.
Additional changes: Inserted Revision History page, restored procedure format
HP/0/B/1009/002 Page 3 of 4 Alternative Method for Determining Dose Rate within the Reactor Building Reference Use
- 1. Purpose This procedure describes an alternative method for determining the approximate dose rate within the reactor building in the event the reactor building monitors (1EMF51A & B or 2EMF51A & B) are inoperable.
The level for this procedure is Reference Use. This procedure must be at the job site at all times.
- 2. References None
- 3. Limits and Precautions 3.1 This procedure is written for use under abnormal conditions which could involve extremely high radiation levels. Only the Radiation Protection Manager (RPM) or designee shall authorize the use of this procedure when needed.
3.2 Appropriate Surveillance and Control coverage shall be used under the direction of the Operations Support Center (OSC).
- 4. Procedure 4.1 Determination of Dose Rate 4.1.1 4.1.2 Ensure that the high range survey instrument to be used is in calibration and a daily response check has been performed.
Obtain a reading by placing the detector in contact with the exterior center portion of the appropriate unit Spent fuel Building to Upper Containment VE Door.
/
Unit 1 (1200C)
R/hr Unit 2 (1250C)
R/hr
HP/0/B/1009/002 Page 4 of 4 4.1.3 Calculate the reactor building dose rate by use of the following equation:
Rs = 1363 x RH Where:
=
Reactor Building Dose Rate (R/hr)
RH
=
Dose Rate outside Spent Fuel Building to Upper Containment VE Door (R/hr) 4.1.4 Forward the following information to the Radiation Protection Technician in the Shift Lab (ext. 4282), TSC Dose Assessors (ext. 4976) or EOF Dose Assessors (382-074{ 0745).
4.1.4.l 4.1.4.2 4.1.4.3 4.1.4.4 Dose rate reading from instrument.
Instrument number.
Your name.
RB doserate calculated in Step 4.1.3.
4.2 Record all necessary information and results on Enclosure 5.1.
4.3 The Reactor Building dose rate shall be determined as directed by the Radiation Protection Manager or designee.
- 5. Enclosures 5.1 Reactor Building Dose Rate Data Sheet 5.2 Variables Used in Determining Dose Rate within the Reactor Building
DATE/TIME.1 Reactor Building Dose Rate Data Sheet HP/0/B/1009/002 Page 1 of 1 Reference Use RX BUILDING SURVEYING INSTRUMENT DOSE RATE (R/HR)
X 1363 =
DOSE RA TE (R/HR) TECHNICIAN NUMBER X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
X 1363 =
June 30, 2015 Memorandum to File McGuire Nuclear Station.2 Variables Used in Determining Dose Rate within the Reactor Building Reference Use
Subject:
McGuire Nuclear Station (MNS)
Variables Used in HP/0/B/1009/002 File: MC-752.05 Plant Radiological Monitoring Condition Statement HP/0/B/1009/002.
Page 1 of 13 MNS Procedure HP/0/B/1009/002, Alternative Method for Determining Dose Rate Within the Reactor Building, describes and provides a method to determine dose rates in the Reactor Building during an emergency situation when 1 EMF51 A&B or 2EMF51 A&B are not available. Revision 2 of the procedure was approved on 11/23/1999 and required the user to obtain a dose rate at the 'Upper Personnel Hatch Air Lock Door (R/hr)' described in step 4.1.3. The current perspective is that it would be extremely difficult to open the VE (Annulus Ventilation) Door at the Upper Containment (U/C) access point and doing so would also present a significant personnel safety concern. A preferential location would be the outside of the VE Door from the associated Spent Fuel Building near the U/C access.
The travel path and this location are more easily accessible, exposes the performer to lower dose rates and does not subject the user to unnecessary safety risks. The selected location at the Spent Fueling Building side of the VE door is also directly aligned with the airlock doors to provide practical dose rate information. The new calculated correction factor was derived using the same methodology as the previous factor but is at a farther distance and accounts for additional shielding encountered by the additional VE door.
References (1) A 'MEMORANDUM TO FILE" dated January 23, 1984 from Catawba Nuclear Station (CNS) is included in the CNS Procedure HP/0/B/1009/006, Alternative Method for Determining Dose Rate within the Reactor Building was referenced and agrees with the methodology.
(2) Concepts of Radiation Dosimetry by Kenneth R. Kase and Walter R. Nelson, Stanford Linear Accelerator Center, Stanford University was used as a reference for the equation in deriving the new correction factor.
(3) Radiological Health Handbook, January 1970 Edition, was used as a reference for density and mass attenuation coefficients.
.2 Variables Used in Determining Dose Rate within the Reactor Building HP/0/B/1009/002 Page 2 of 13 The applicable pages from Reference 2 are provided below for documentation.
Below are the diagrams and equations for calculating the dose rate outside containment.
- 5. 8 Right-Circular Cylinder Source: Infinite-Slab Shield, Uniform Activity Distribution I
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1767Al8
- 118 -
.2 Variables Used in Determining Dose Rate within the Reactor Building HP/0/B/1009/002 Page 3 of 13 The slab absorber is parallel to the cylinder a.xis. The source strength per unit volume, Sv,, is constant. The exact solution3 of this problem is very lengthy and is not generally used. What is usually done1 (FBM) is to approximate the cyl-inder by a line source of strength SL= rr R~8v which is positioned within the cylinder to correctly account for self absorption. There is no simple expression for Z = Z(R0, a, b). the self absorption distance; however, by empirically fitting the approximate method to the exact calculations, 3 only three curves for Z plus the F(6, b) curves (that is, the Sievert integrals) for line sources are needed in order to solve cylinder-slab problems. The three curves needed to obtain Z are given in the Appendix and are used as follows; CASE:
a/R0 ~ 10 Use figure A.20 (see Appendix) andµ 8 R0 to obtainµ 8 Z, whereµ 8(cm-1) is the macroscopic source attenuation coefficient. Then obtain b2 from where b1=~µ..t.
i 1 1 Finally, obtain the flux density at P 1 from Sv,R~
<P1 = 4(a+ Z) [F(Ol' ba) + F{82* b2>J and at P 2 from (5.27)
(5.28)
(5.29)
(5. 30) using the F-functions which are plotted in the Appendix. These estimates of the flux density are supposedly good to ;:1: 10%,
- provided a/R0 2::. 10.
- Note: Provided that the correct buildup factors have been included.
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CASE:
a/Ro < 10.2 Variables Used in Determining Dose Rate within the Reactor Building HP/0/B/1009/002 Page 4 of 13 Use Figs. A. Zl and A. 22 in conjunction with each other to obtainµ z. That s
is, knowing n0, a, and µ 8, find m from. the first graph; knowing a/R0 and b1, fin
µ.
8 Z/m from the second graph; then multiply these together to obtainµ 6 Z. Fina.ll~
follow the recipe above to obtain <j>. This approximation will be good to+ 40% a.nc
-5%.
Other formulas are given for cylinders viewed exterior on end, and interior1 (FBM).
Example:
Consider a cylindrical tank containing radioactive water uniformly distribute1 throughout. The field positron is P 1 with 01 = 82, and the distance is restricted 1 R 0
~ a $ 70. O inches with
- 1)
R 0 = 5. 5 inches
- 2) h = 14. 0 inches
- 3) no shielding or buildup
- 4) self absorption in the water and
- 5) the radioactive source consists mainly of O. 511 MeV photons with
µ
= 0. 092. cm-l (the total attenuation coefficient for water).
s The normalized flux density is obtained from Eq. (5. 29), and is where we have dropped the subscript on theta, and where tan (J== h/2(a + Z) b2== µSZ
- Using Figs. A.14, 20, 21, and 22, we obtain Table 5.1.
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.2 Variables Used in Determining Dose Rate within the Reactor Building HP/0/B/1009/002 Page 5 of 13 Below are the diagrams and equations for calculating the dose rate inside containment.
5.10 Spherical Source: Field Position at Center of Sphere z
y X
1767A7'
- 125 -
-1r r
_,, r rs rs 2
8ve dV ~e r sin9 dr d8 dP d<p =
2
=
4~r 4~r2 (5. 32)
This submersion situation is applicable to finding the dose rate in a radioactive cloud o:r in a body of contaminated water. By symmetry, the uncollided flux from a hemi-sphere (that is, no buildup) is exactly one-ha.Jf of this.
.2 Variables Used in Determining Dose Rate within the Reactor Building HP/0/B/1009/002 Page 6 of 13 Provided on the next three (3) pages are the figures used from Reference 2 in solving equation 5.29.
Specifically; Fig A.21 is used to determine m Fig A.22 is used to determine ~µ,8 Z m
Fig A.14 is used to determine F(l11,b2)andF(flz,h2 ).
2.6 2.. 4 2.0 1.6 E
1.2 0.8 0.4 0
0.2 Variables Used in Determining Dose Rate within the Reactor Building 2
fLs (a+ Ro) = 20.0 4
a/Ro FIG. A.21 9.0 8.0 6.0 4.0
- 2.0 6
HP/0/B/1009/002 Page 7 of 13 18.. 0 10.0 8.5 7.0 5.0 3.0 1.0 8
1767... 20 self-absorption distance, Z, of a cylinder for a/R0 < 10.
Note: Use in conjunction with Fig. A.22
_ 187 _
_J
N Cf)
- L.
--E 4.4 4.0 3.6 3.2 2.s*
2.4 2.0 1.6 1.2 0.8 0.4 0
0.2 Variables Used in Determining Dose Rate within the Reactor Building 4
8 12 b
FIG. A. 22 16 20 self-absorption distance, z. of a cylinder for a/R0 < 10.
Note: Use in conjunction with Fig. A. 21
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HP/0/B/1009/002 Page 8 of 13 24 1767A21
100 10-I
..0..
CJ:> -
u..
,0-2 0.2 Variables Used in Determining Dose Rate within the Reactor Building 90° 60° 40° 30° 20° 2
3 4
5 b
FIG. A.14
- 180 -
6 HP/0/B/1009/002 Page 9 of 13 7
1767A.22
Calculation and Solution Assumptions:.2 Variables Used in Determining Dose Rate within the Reactor Building The average gamma energy per disintegration is 0.5 MeV
- 1 photon per decay HP/0/B/1009/002 Page 10 of 13 Steel (VE door and airlock doors) was modeled as iron (Fe) when selecting density and mass attenuation coefficients Radioactive concentrations inside containment were considered an instantaneous uniform concentration Assumed the concentration in Upper Containment was the only influence on dose rate calculated at SFP VE door STP was assumed for the density and mass attenuation coefficient of air Dose rates inside containment were calculated assuming a sphere of infinite radioactive cloud of radius Ro =57'6".
MNS Model Drawing:
Distribution
/
/
' ' '/
h
/'
/,
/
/
/ ---
I 82 I
I I La 1-- t-1 Where:
Ra=57' 6" h = 64' 3" and is considered from the U/C Floor to Dome a= 22' t=3.5" 81=32° 82=2° Z=49.9' e,
P2
~
/ 82
-.... P1 I
l I
1 I
I
.1 l707Al8 P1=detector location (on SFP side of VE Door at SFP Door at U/C)
.2 Variables Used in Determining Dose Rate within the Reactor Building HP/0/B/1009/002 Page 11 of 13 Using equation 5.29 of Reference 2, and following the steps outlined for CASE: a/Ro<10 on page 120, the flux outside containment at point P1 is calculated as follows:
..::. =
22
' =.383 Ro 57.51 cm2 fl
µs =.. 0870 -
X.001293 -
3 = 1.12 X 10-4 cm-1 fl cm Note: the mass attenuation coefficient and density used to find µ 3 were taken from Reference 3.
h 1 = L µ;;ti=, (661.67 cm)(1.12 X 10-4 cm-1-) + (8.89cm)(.471cm.-1 ) = 4.26 i
From figure A.21, m= 0.1, using a/Ro and µs(a+Ro)
From figure A.22 since m = O.l.
1 -µ~=2.. 5 2n and z = 2232 cm= 73.2' and From figure A.14 F(3Z0,4.S) = 4 X i,0-3 and F(2°,*4.5) =>>4X 10-t
.2 Variables Used in Determining Dose Rate within the Reactor Building
=
S1.1(l?S2.6 cm) 2
[(4X 10-3 ) + (4X 10-4 )] = 1.16.S
(/J 4(670.56,cm-!l-2232 cm) v Note: (fJ is the flux is at the detector or A1 and Sv is the activity HP/0/B/1009/002 Page 12 of 13 Using equation 5.32 of Reference 2, the flux inside containment is calculated as follows:
Finally the ratio of the inside and outside containment flux are given as:
1591.3.s'lf
---=1363 1.168 S17 Note: sv (activity) is constant for both equations and cancels on division.
Conclusion.2 Variables Used in Determining Dose Rate within the Reactor Building HP/0/B/1009/002 Page 13 of 13 Applying the same methodology as previously used but at an increased distance and accounting for additional shielding results in the following equation for use in MNS Procedure HP/0/B/1009/002:
DRi = 1363 X.DRo Where:
DRi=Dose Rate Inside the Reactor Building (indicative* of Unit 1/2 EMF51 A&B Dose Rate Reading)
DRo=Dose Rate obtained on SFP side of SFP VE Door at Upper Containment This number is applicable to either unit at McGuire Nuclear Station and is independent of concentration or time after shutdown.
Prepared By: Cody Breitkreuz Date: 06/15/2015 Reviewed By: Chris Whitener Date: 06/30/2015