ML20216F883
| ML20216F883 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 04/13/1998 |
| From: | Zinke G Maine Yankee |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20216F888 | List: |
| References | |
| GAZ-98-21, MN-98-25, NUDOCS 9804170270 | |
| Download: ML20216F883 (8) | |
Text
MaineYankee
, , P.O. BOX 408
- WISCASSET, MAINE 04578 * (207) 882-6321
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1 April 13, 1998 MN-98 25 GAZ-98-21 Proposed Change No. 208 j UNITED STATES NUCLEAR REGULATORY COMMISSION Attention: Document Control Desk i Washington, DC 20555 -
Reference:
(a) License No. DPR-36 (Docket No. 50-309)
(b) Letter: M. J. Meisner to USNRC: Proposed Technical Specification '
Change No. 207 - Permanently Defueled Technical Specifications:
MN-97-116, dated October 20, 1997 (c) Letter: USNRC to M. J. Meisner: Issuance of Amendment No. 161 to Facility Operating License No. DPR-36 Maine Yankee Atomic Power Company (TAC No. M99862)
Subject:
Proposed Technical Specification Change No. 208 - Fuel Storage Pool Water Level:
l Gentlemen:
In Reference (b), Maine Yankee submitted a proposed change to the Technical Specifications to reflect the permanently defueled status of the plant. That 1 proposed change was based upon existing accident analysis for Maine Yankee and.
among other things specified the Fuel Storage Pool Water Level at 23 feet over the top of irradiated fuel assemblies seated in the storage racks during movement of ,
irradiated fuel assemblies in the fuel storage pool. NRC approved this proposed )
change in Reference (c). '
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l Maine Yankee hereby submits, pursuant to 10 CFR 50.90, an application to amend the I Technical Specifications to base the Limiting Condition for Operation for the Fuel Storage Pool Water Level on the revised analysis for the Fuel Handling Accident and on a new analysis for radiological shielding during movement of irradiated fuel.
These analyses are effective as of December 6, 1997', one-year after the last shutdown from reactor operations.
Attachmei t I to this letter provides a description of the revised analysis, the proposed changes for the LC0 for Fuel Storage Pool Water Level and the Significant Hazards Evaluation. Attachment 11 to this letter provides a copy of the proposed Technical Specification 3.1.1. Attachment III contains the proposed Bases for the TS 3.1.1.
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9804170270 900413 PDR ADOCK 05000309
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MaineYankee l UNITED STATES NUCLEAR REGULATORY COMMISSION MN-98-25 Attention:, Document Control Desk Page Two This change does not involve a significant increase in the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a l significant reduction in the margin of safety. Based on our evaluation, we conclude l there is reasonable assurance that the Maine Yankee plant activities, consistent l with the proposed Technical Specifications. will not impact the health and safety of
- the public.
l This proposed change has been reviewed and approved by the Plant Operation Review Committee. The Nuclear Safety Audit and Review Committee and the Independent Review and Audit Committee have also reviewed this submittal. A representative of the State of Maine is being informed of this request by a copy of this letter.
If you have any questions, please contact us.
Very truly yours, l George A. Zinke. Director l Nuclear Safety & Regulatory Affairs l
! Attachments c: Mr. Hubert Miller Mr. Michael Webb i Mr. Singh Bajwa l
Mr. R. A. Rasmussen i Mr. Clough Toppan l Mr. Patrick J. Dostie Mr. Uldis Vanags STATE OF MAINE Then personally appeared before me. George A. Zinke, who being duly sworn did state that he is the Director, Nuclear Safety & Regulatory Affairs of Maine Yankee Atomic l Power Company, that he is duly authorized to execute and file the foregoing request
, in the name and on the behalf of Maine Yankee Atomic Power Com;;any, and that the l l statements therein are true to the best of his knowledge and belief.
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&d.,. .
Notary Public Monica W. Fortier, Notary Pubiki US Nuclear Regulatory Commission State of McIne Attn: Document Control Desk My Commission Expiron 5/3/09 Washington, DC 20555 l
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f 1 l ATTACHMENT I Page 1 of 5
{ , , DESCRIPTION OF CHANGES PROPOSED CHANGE NO. 208 BACKGROUEQ l On October 20, 1997 Maine Yankee submitted Proposed Change No. 207 (Reference b).
The purpose of that submittal was to propose revised technical specifications appropriate for the permanently shutdown and defueled status of the facility. The NRC approved that proposed change in Reference (c). That proposed change was based upon existing analysis previously completed for Maine Yankee.
The previously existing accident analysis for the fuel handling accident assumed two i cases. The first case is the realistic case where the accident is assumed to occur '
one week after shutdown from operations and a decontamination factor for iodine of l 500. The second case is the conservative case where the accident is assumed to j occur seventy-two hours after shutdown from operations and a decontamination factor of 133 and 1 for inorganic lodine and organic iodine. For the purpose of establishing an upper limit on the amount of fuel damage resulting from a fuel handling accident, it is assumed that the fuel assembly is dropped during handling.
The number of ruptured fuel rods which would result depends on several variables including the kinetic energy at impact and fuel assembly orientation during impact.
The existing analyses indicate that if a fuel assembly were dropped to the bottom of the spent fuel pool and then rotated and struck a protruding structure, only the outer row of fuel rods would fail. However, to assure that the limiting case is considered. it is assumed that all rods in the dropped assembly fail upon impact.
The resulting doses which are calculated for the fuel handling accident are below the values specified in 10CFR100. As described in the basis of the Technical Specification (formerly specified as TS 3.13), the analysis of the fuel handling accident conducted by the NRC assumed 23 feet of water and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay time.
The NRC's analysis resulted in dose consequences limited to 10% of 10CFR100.
A revised analysis has been completed for the Fuel Handling Accident in the Spent Fuel Pool. The revised analysis is based on the Fuel Handling Accident described in the FSAR with the following changes:
l The fuel rod gap inventory for release is based upon the worst case (highest burnup, enrichment and longest operating history) fuel l
assembly presently stored in the spent fuel pool .
The doses are evaluated as a function of time after shutdown (starting one year after the 12/6/96 shutdown) at the Exclusion Area Boundary.
The doses are evaluated at the Control Room intakes and inside the Control Room with various ventilation system operating conditions.
Reduced iodine Decontamination Factors (DF's) are used based on two cases: (a) DF - 1. No radionuclide retention within the SFP: and (b) DF - 75. SFP Water level of 19 feet.
. The fuel assembly that is postulated to be dropped, releases 100% of I
the assembly maximum gap radionuclide inventory of Kr-85,1-129. I-131.
Xe-131m and Xe-133.
. Assembly gap radionuclide inventories consist of 10% core iodine's (except I-129 - 30%) and 10% core noble gases (except Kr85 - 30%).
These gap inventories are releasea to the pool water with a subsequent t release of 100% of the noble gases.
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ATTACHMENT I Page 2 of 5 QESCRIPTION OF CHANGES PROPOSED CHANGE NO. 208 e . .
. No fuel overheating occurs and, therefore no significant solid fission products are released.
. No credit is taken for isolation of the Spent Fuel Building. The release is modeled as an instantaneous puff release using appropriate diffusion factors.
l . No credit is assumed for Control Room isolation or air filtration by
! either the Control Room intake air system or the Control Room l recirculation system.
! . The radionuclide inventories (Ci) for the maximum assembly contained within the spent fuel pool (as of 12/6/97) and used to assess both offsite and Control Room doses are as follows:
Kr-85 - 4.04E+03 1-129 - 1.85E-02 I-131 - 6.60E-09 Xe-131m - 5.86E-06 Xe-133 - 8.06E-16
, The revised analysis, effective one year after the 12/6/96 shutdown, shows that for i a DF-75 (SFP level of 19 feet) the Fuel Handling Accident would result in the l
following dose at the Exclusion Area Boundary (EAB):
EAB Dose (milliren0 EAR Part 100 Limit EPA PAG Limit Thyroid: 0.195 300.000 5.000 Effective Whole Body: 0.224 25,000 1.000 Skin: 31.7 N/A 50.000 The results show that the projected doses are insignificant in comparison to 10 CFR Part 100 limits and less than the Environmental Protection Agency (EPA) Protective Action Guideline (PAG) values.
The revised analysis, effective one year after the 12/6/96 shutdown, also shows that for the.above conditions the Fuel Handling Accident would result in the following dose at the Control Room (CR) Intakes:
CR Intake Dose (millirem) CR Intakes GDC 19 Limit l Thyroid: 4.3 30.000 Effective Whole Body: 7.2 5.000 Skin: 350 N/A The results show that the calculated doses at the Control Room intakes are within 10 CFR Part 50. General Design Criterion (GDC) 19 dose limits. !
l The margin of safety for the fuel handling accident relates to the acceptance limit which the NRC had accepted during its review of the license. The fuel handling i accident acceptance limit defined in the basis for the Maine Yankee Technical '
Specification (formerly specified as TS 3.13.D.10) is 10% of 10 CFR Part 100 limits.
A reduction in margin of safety occurs when the acceptance limit is no longer met as a result of a proposed change, test or experiment, i'he margin that exists between the technical specification limit for the fuel storage pool water level and the fuel handling accident acceptance limit represents operating margin. {
s , ATTACHMENT f Page 2 of 5 DESCRIPTION OF CHANGES PROPOSED CHANGE NO. 208 No fuel overheating occurs and, therefore, no significant solid fission products are released.
No credit is taken for isolation of the Spent Fuel Building. The release is modeled as an instantaneous puff release using appropriate diffusion factors.
No credit is assumed for Control Room isolation or air filtration by either the Control Room intake air system or the Control Room recirculation system.
The radionuc11de inventories (Ci) for the maximum assembly contained
- within the spent fuel pool (as of 12/6/97) and used to assess both offsite and Control Room doses are as follows
Kr-85 - 4.04E+03 I-129 - 1.85E-02 I-131 - 6.60E-09 Xe-131m - 5.86E-06 Xe-133 - 8.06E-16 The revised analysis, effective one year after the 12/6/96 shutdown, shows that for a DF-75 (SFP level of 19 feet) the Fuel Handling Accident would result in the following dose at the Exclusion Area Boundary (EAB):
EAB Dose (millirem) FR Part 100 Limit EPA PAG Limit )
Thyroid: 0.195 300,000 5,000 l l Effective Whole Body: 0.224 25,030 1,000 Skin: 31.7 N/A 50,000 {
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The results show that the projected doses are insign oilcant in comparison to 10 CFR Part 100 limits and less than the Environmental Protection Agency (EPA) Protective l
Action Guideline (PAG) values. ,
The revised analysis, effective one year after the 12/6/96 shutdown, also shows that for the above conditions the Fuel Handling Accident would result in the following !
dose at the Control Room (CR) Intakes:
CR Intake Dose (millirem) fB Intakes GDC 19 Limit l Thyroid: 4.3 30,000 l Effective Whole Body: 7.2 5.000 Skin: 350 N/A The results show that the calculated doses at the Control Room intakes are within 10 j CFR Part 50 General Design Criterlon (GDC) 19 dose limits.
The margin of safety for the fuel handling accident relates to the acceptance limit which the NRC had accepted during its review of the license. The fuel handling accident acceptance limit defined in the basis for the Maine Yankee Technical Specification (formerly specified as TS 3.13.D.10) is 10% of 10 CFR Part 100 limits.
A reduction in margin of safety occurs when the acceptance limit is no Jonger met as a result of a proposed change, test or experiment. The margin that exists between the technical specification limit for the fuel storage pool water level and the fuel handling accident acceptance limit represents operating margin.
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. ATTACHMENT I Page 3 of 5 DESCRIPTION OF CHANGES i PROPOSED CHANGE NO. 208 j - '
l A radiological shielding analysis was performed to determine minimum water level above the top of a fuel assembly (bottom of the flow plate) to maintain the ;
radiation dose rates less than 50 mrem /hr at the fuel handling crane platform and '
l the walkway around the pool. This analysis determined that five-and-one-half feet l of water above the top of a fuel assembly (bottom of the flow plate) would provide l sufficient shielding to meet this objective. This water level will result in a l radiation dose rate at the surface of the water over a raised fuel assembly which is less than 80 mrem /hr. The margin of safety for the radiological shielding consideration relates to the acceptance limit which the NRC had accepted during its review of the license. The NRC acceptance limit defined in the Maine Yankee FSAR is l that combination of occupancy time and dose rate such that no station personnel l receive in excess of 5 rem per year (10 CFR 20.1201). A reduction in margin of safety occurs when the acceptance limit is no longer met as a result of a proposed l change. test or experiment.
DESCRIPTION OF CHANGES TS 3.1.1 Spent Fuel Pool Water Level - One of the two LC0's and associated surveillance requirements, contained in the TS is for the Spent Fuel Pool Water Level. This specification was determined to be necessary to meet 10CFR50.36(c)(2)(ii)(B) Criterion 2, in that it is a limit that is an initial condition for a design basis accident, namely the fuel handling accident.
l l This specification requires that the spent fuel pool water level be maintained within limits during fuel movement and requires verification of compliance at a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> frequency. The actions associated with this LC0 require immediate suspension of movement of irradiated fuel assemblies in the fuel storage pool. The immediate actions minimize the potential for a fuel handling accident or a loss of shielding during fuel movement. Restrictions on crane operation with loads over the spent fuel pool are not included in the LCO, consistent with the improved Standard Technical Specification. These restrictions are described in the FSAR.
l The minimum water level in the fuel storage pool meets the assumptions of iodine l decontamination factors following a fuel handling accident. Reference (b) proposed l that this minimum water level be maintained at 23 feet above the top of a fuel
- assembly seated in the storage racks during movement of irradiated fuel assemblies l in the fuel storage pool. This proposed change is proposing to revise this minimum water level to be maintained at 21 feet above the top of a fuel assembly seated in the storage racks during movement of irradiated fuel assemblies in the fuel storage pool. As in Reference (b), this proposed change defines in the basis that the top of the fuel assembly is the bottom of the flow plate. The revised analysis was l performed assuming a fuel handling accident occurs after the spent fuel has decayed for at least one year. The initial conditions assumed a minimum of 19 feet of water l
i for iodine absorption. No credit was taken for control room or spent fuel pool ventilation filtration or isolation. The results of the revised analysis demonstrate that the projected EAB doses resulting from a postulated fuel handling accident are insignificant in comparison to 10 CFR Part 100 limits. The specified water level also provides more than five-and-one-half feet of water shielding over a raised fuel assembly assumed to be at the maximum allowable height of eighteen inches above the top of the lowest fuel storage rack. The shielding resulting from this water level
, ATTACHMENT I Page 4 of 5 t
DESCRfPTION OF CHANGES PROPOSED CHANGE NO. 208
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maintains the radiation dose rate less than 50 mrem /hr at the fuel handling hoist
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I platform or the walkway around the pool and less than 80 mrem /hr at the surface of l the water )
l l SIGNIFICANT HAZARDS EVALUATION i
The proposed change to the Technical Specifications, has been evaluated against the i standards of 10 CFR 50.92 and has been determined not to involve a significant l l hazards consideration. An evaluation against these standards is provided below:
TS 3.1.1 Spant Fuel Pool Water Level The proposed change does not:
- 1. Involve a significant increase in the probability or consequence of an accident previously evaluated.
l The proposed restrictions on the water level in the spent fuel pool has no )
impact on the probability or consequences of the remaining applicable l design basis accidents. These restrictions are fulfilled by normal operating conditions, preserve initial conditions assumed in the analyses of postulated DBAs and ensure that the conditions of such DBAs are consistent with the analyses. Revised analysis was performed assuming a fuel handling accident occurs after the spent fuel fission products have decayed at least one year The initial conditions assumed a minimum of 19 feet of water for 1odine absorption. No credit was taken for control room or spent fuel pool ventilation filtration. The results of the revised analysis demonstrate that the projected doses resulting from a postulated fuel handling accident are insignificant in comparison to 10 CFR Part 100 limits. Therefore, the proposed changes to the Technical Specifications do not involve any increase in the probability or consequences of any accident previously evaluated.
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- 2. Create the possibility of a new or different kind of accident from any accident previously evaluated.
The proposed restrictions on the water level in the spent fuel pool are fulfilled by normal operating conditions and preserve initial conditions i
assumed in the analyses of postulated DBAS. These additional restrictions I do not involve changes to any structure or 'quipment affecting the safe l storage cf irradiated fuel. The results of toa revised analysis of a fuel handling accident demonstrate that the projected doses are insignificant in comparison to 10 CFR Part 100 limits with a minimum of 19 feet of water for iodine absorption. In addition, maintaining this minimum water level will also provided sufficient shielding for personnel radiation protection during fuel movement. Therefore, the proposed changes to the Technical Specifications would not create the possibility of a new or different kind of accident from any accident previously evaluated.
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l , ATTACHMENT I Page 5 of 5 l DESCRIPTION OF CHANGES PROPOSED CHANGE NO. 208 l .
- 3. Involve a significant reduction in a margin of safety.
The proposed restrictions on the water level in the spent fuel pool preserve initial conditions assumed in the analyses of postulated DBAs and ensure that margins of safety contained in the analyses are maintained.
The margin of safety for the fuel handling accident relates to the acceptance limit which the NRC approved during its review of the license.
The fuel handling accident acceptance limit defined in the basis for the l Maine Yankee Technical Specification (formerly specified as TS 3.13.D.10) is 10% of 10 CFR Part 100 limits. A reduction in margin of safety occurs when the acceptance limit would no longer be met as a result of a proposed change. Since the acceptance limit is met, there is no reduction in margin of safety. The projected dose rates at the specified Fuel Storage Pool water level during fuel movement with a fuel assembly raised to its
- highest allowable height would result in personnel exposures within that previously assumed. There is no reduction in a margin of safety. The NRC acceptance limit which is that combination of occupancy time and dose rate i that maintains personnel doses within 10 CFR 20.1201 limits is not I exceeded. Therefore, the proposed changes to the MYTS would not involve a significant reduction in any margin of safety.
Conclusion Maine Yankee has concluded that the proposed change to the Technical Specifications does not involve a significant hazards consideration as defined by 10 CFR 50.92.
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