ML20086Q817
| ML20086Q817 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 07/26/1995 |
| From: | Barkhurst R ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20086Q821 | List: |
| References | |
| W3F1-95-0017, W3F1-95-17, NUDOCS 9507280203 | |
| Download: ML20086Q817 (15) | |
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Khym ! A TOM /i 0761 k: 504 739 fM1 Ross P. Barkhurst
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A4.05 PR July 26, 1995 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C.
20555 I
Subject:
Waterford 3 SES Docket No. 50-382 1
License No. NPF-38 l
Technical Specification Change Request NPF-38-167
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Gentlemen:
The attached description and safety analysis supports a change to the Waterford 3 Technical Specifications (TS).
The proposed change incorporates provisions into TS 3.9.4 that will allow l
the containment personnel airlock and equipment doors in the containment building, to be open under strict administrative controls during core alterations or movement of irradiated fuel. This request will also allow containment isolation valves to be open on an intermittent basis during core alterations or movement of irradiated fuel in the containment.
This proposed change demonstrates compliance with the acceptance criteria that forms the basis for TS 3.9.4 and will continue to preserve the assumptions of the safety analysis. However, this change will have a dramatic affect on refueling outages by permitting increased efficiency.
1 Therefore, this proposed change is being submitted as-part of the Cost Beneficial Licensing Action (CBLA) program established within NRR where increased priority is granted to licensee requests for changes requiring
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staff-review that involve high cost without a commensurate safety benefit.
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Technical Specification Change Request NPF-38-167 W3F1-95-0017 Page 2 July 26, 1995 The proposed change was developed to decrease the operational burden placed on outage resources.
We expect the proposed change will realize safety benefits and a cost reduction of approximately $14,988,000 over the life of the plant.
The proposed change has been evaluated in accordance with 10CFR50.91(a)(1) using criteria in 10CFR50.92(c) and it has been determined that the proposed change involves no significant hazards considerations. The Plant Operations Review and Safety Review Committees have reviewed and accepted the proposed change.
The circumstances surrounding this change do not meet the NRC's criteria for exigent or emergency review.
However, due to the impact on our refueling outages, we respectfully request an expeditious review pursuant to the CBLA program.
Should you have any questions or comments concerning this request, please contact Paul Caropino at (504) 739-6692.
Very truly yours, R.P. Barkhurst Vice President, Operations Waterford 3 RPB/PLC/tmm
Attachment:
Affidavit NPF-38-167 cc:
L.J. Callan (NRC Region IV), C.P. Patel (NRC-NRR),
i R.B. McGehee, N.S. Reynolds, NRC Resident Inspectors Office, Administrator Radiation Protection Division (State of Louisiana), American Nuclear Insurers
j UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
'In the matter of
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Entergy Operations, Incorporated
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Docket No. 50-382 Waterford 3 Steam Electric Station
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i AFFIDAVIT R.P. Barkhurst, being duly sworn, hereby deposes and says that he is Vice President Operations - Waterford 3 of Entergy Operations, Incorporated; that he is duly authorized to sign and file with the Nuclear Regulatory Commission the attached Technical Specification Change Request NPF-38-167; that he is familiar with the content thereof; and that the matters set forth therein are true and correct to the best of iiis knowledge, information and belief.
ML M
R.P. Barkhurst Vice Presider 4t Operations - Waterford 3 STATE OF LOUISIANA
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) ss PARISH OF ST. CHARLES
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Subscribed and sworn to before me, a Notary Public in and for the Parish and State above named this 2.C" day of Jus
, 1995.
S.
Notary Public My Commission expires wen '-' F t r
I DESCRIPTION AND SAFETY ANALYSIS OF PROPOSED CHANGE NPF-38-167 This proposed change revises Technical Specification (TS) 3.9.4 and its associated bases to incorporate provisions that will allow the containment personnel airlock and/or equipment hatch doors, and other penetrations to be open under strict administrative controls during core alterations or movement of irradiated fuel in the containment building.
Existino Specification See Attachment A Proposed Specification See Attachment B Backaround Limiting Condition for Operation (LCO) 3.9.4, Containment Building Penetrations, is intended to prohibit a release of fission product radioactivity to the environment due to a Fuel Handling Accident.
The LC0 requires the containment equipment door, the containment airlocks, and containment penetrations that provide direct access from containment atmosphere to outside atmosphere (excluding containment purge) to be closed during core alterations or movement of irradiated fuel in the containment.
Penetrations for containment purge must be capable of being closed by an automatic containment purge valve.
In MODE 6, the potential for containment pressurization as a result of an accident is not likely.
Therefore, requirements to isolate the containment from the outside atmosphere are less stringent.
LC0 3.9.4 specifies requirements that facilitate " containment closure" rather than " CONTAINMENT INTEGRITY." CONTAINMENT INTEGRITY is required during operational Modes 1 through 4 and preserved by other LCOs that specify more stringent requirements (e.g., isolation and leak testing).
Containment closure means that all potential escape paths are closed or capable of being closed. Since there is no potential for containment pressurization, the Appendix J leakage criteria and tests are not required.
r During core alterations or movement of irradiated fuel assemblies within containment, the most severe radiological consequences result from a Fuel Handling Accident (FHA).
The FHA is a postulated event that involves damage to irradiated fuel.
FHAs, analyzed in the Final Safety Analysis Report (FSAR), include dropping a single irradiated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies. The FHA analysis assumes that there is 23 feet of water above the fuel and that the fuel has been subject to a minimum decay time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The TS include these restrictions prior to core alterations to ensure that the release of fission product radioactivity, subsequent to a FHA, results in doses that are well within the guideline values specified in 10CFR100. The acceptance limits for offsite radiation exposure are contained in Standard Review Plan (SRP) Section 15.7.4, which defines "well within" 10CFR100 as 25% or less of the 10CFR100 values.
The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10CFR100.
The LC0 limits the consequences of an FHA in containment by limiting the potential escape paths for fission product radioactivity released within containment. The proposed change will continue to meet the intent of LCO 3.9.4 by ensuring that escape paths are closed or capable of being closed in a rapid manner such that the acceptance limits for offsite radiation exposure are met.
Description The containment equipment door is a welded steel assembly, with a double gasketed flanged and bolted cover.
This equipment access is 14 feet in diameter and provides a means for moving large equipment and components into and out of containment during refueling outages.
TS 3.9.4.a currently requires the equipment door to be closed and held in place by at least four bolts during core alterations or movement of irradiated fuel assemblies within containment.
The proposed change will permit the equipment door to be open provided that the door is capable of being immediately closed.
In addition, trained and qualified individuals shall be available to close the door if called upon. These individuals are trained in performing equipment door closure.
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.q On September 29, 1989, as part of the initiatives to address Generic Letter 88-17, Waterford 3 performed an equipment hatch closure test to verify the time required to close the equipment hatch. The test simulated conditions normally found during an outage. The. total time required to close the equipment hatch was fourteen (14) minutes and twenty five (25) seconds.
Measures for the expeditious closure of the equipment hatch were established to address Generic Letter 88-17, and the measures are documented in procedures MM-008-001 and OP-001-003.
The containment personnel airlock (PAL) connects the containment interior with the Auxiliary Building. The PAL is provided for the purpose of permitting personnel to enter and exit the containment. The PAL contains-two airlock doors with a personnel chamber between the doors. TS 3.9.4.b currently requires a minimum of one door in each airlock to be closed during core alterations or movement of irradiated fuel within the containment.
The proposed change would permit the PAL doors to be open provided that at least one door is capable of closing and
- iesignated individual is available to close the PAL door. The containmen; s also provided with an Emergency Airlock (EAL) in addition to the PAL. The EAL is a smaller airlock which connects the containment with the outside environs. This change does not affect the EAL.
Technical Specification 3.9.4.c currently requires that during core alterations or movement of irradiated fuel within containment each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either: (1) closed by an isolation valve, blind flange, or manual valve, or (2) be capable of being closed by an operable containment purge valve. During the performance of Local Leak Rate Testing certain containment isolation valves (i.e., those subject to Type C testing) are required to be opened in order to drain the penetration piping, providing direct access from the containment atmosphere to the outside atmosphere.
Accordingly, LLRT tests cannot be performed during core alterations or fuel movement inside containment. This restriction significantly complicates the logistics for performing LLRT and reduces overall refuel efficiencies.
The i
proposed change to TS 3.9.4.c would allow containment penetrations to be open, on an intermittent basis under administrative control provided that the penetration is capable of being closed by an isolation valve. The administrative controls are the same controls that.are applicable during operational Modes 1 through 4 (TS 3/4.6.3) and are included in the Bases. The opening of containment isolation valves on a intermittent basis under administrative control includes the following considerations: (1) stationing i
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.c an operator /STA, who is in constant communication with control room, at the valve controls, (2) instructing this operator /STA to close these valves in an F
accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.
The proposed change will allow containment penetrations to be open, as described above, when the plant is in Mode 6 with 23 feet of water above the fuel. This is an assumption used in the FHA analysis and, therefore, included in the TS requirements.
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A minor change to TS Surveillance Requirement 4.9.4 was necessary to support I
the proposed change described above. This is purely an administrative change.
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As previously stated the most severe radiological consequences during core N
alterations or movement of irradiated fuel assemblies within containment, result from an FHA. Waterford 3 has evaluated the consequences of an FHA in the Reactor Containment Building. The proposed change is based on the conservatisms that are included in the FHA analysis and the fact that the analysis results continue to be well within the acceptance limits.
The evaluation for the offsite and control room radiological consequences of an FHA in the reactor containment building used the TRANSACT computer code an enhancement of the NRC approved TAC V code.
The assumptions used in the calculations are in accordance with the RG 1.25 recommendations.
The analysis also considered the following conservative assumptions in adddition to RG 1.25:
Number of Failed Fuel Rods:
The number of failed fuel rods (i.e., 60 failed fuel rods) is based on the worst postulated assembly drop.
Gap Inventory:
For each isotope the most conservative calculated activity (i.e., the highest value) for the 60, 40 and 20 GWD/MT burnups is assumed.
Release Duration: An instantaneous release of the containment activities to the outside atmosphere is assumed.
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The total number of failed fuel rods are based on the Waterford 3 design basis fuel handling accident in the Fuel Handling Building described in the Waterford 3 FSAR Section 15.7.3.4.
This design basis analysis establishes that the worst case Fuel Handling Building accident is the failure of fuel rods in four rows parallel to one assembly face i.e., 60 fuel rods. This analysis employs the conservative assumption that the dropped fuel assembly at impact has reached its terminal velocity in water. The analysis assumes that all of the kinetic energy of the fuel assembly at impact is absorbed only by the fuel rods at a single line of contact. With this assumption, no more than four rows, 60 fuel rods, will undergo failure.
Since the fuel assembly is l
travelling at its terminal velocity in water at the time of impact, the number of failed fuel rods is independent of the distance through which the fuel assembly is assumed to drop. Therefore, the Fuel Handling Building accident 1
analysis is a bounding analysis for any hypothetical fuel assembly drop inside containment.
Regulatory Position 1.d of RG 1.25 states that all of the gap activity in the damaged rods is released and consists of 10% of the total noble gases other than Kr-85, 30% of the Kr-85, and 10% of the total radioactive iodine in the rods at the time of the accident.
The Waterford 3 gap inventories were
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obtained from a calculation performed by Entergy Central Design Engineering.
The assumptions used in generating the fuel rods gap inventories are consistent with RG 1.25.
The gap inventories calculated by Entergy are more conservative than the gap inventories currently used in the FSAR.
The gap activity was calculated for the 60, 40 and 20 GWD/MT burnups.
For conservatism, for each isotope the highest activity for the 60, 40 and 20 GWD/MT was assumed.
Regulatory Position 1.i of RG 1.25 states that the radioactive material escapes from the pool to the building is released from the building over a two hour time period. The Waterford 3 calculation assumes that the noble gases and radiciodine from the gap of the broken fuel rods are instantaneously released to the containment atmosphere.
Further, all the airborne radioactivity reaching the containment is assumed to be released instantaneously to the outside atmosphere. This assumption is overly conservative, since for all practical purposes, it ignores the existence of the containment building.
In an actual event the personnel airlock and equipsit hatch doors will be closed as soon as all personnel inside containment are evacuated.
The calculated 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> offsite and 30 day control room dose consequences (rem) are:
Dose Point Thyroid Whole Body Exclusion Area Boundary 39.950 0.1116 Low Population Zone Boundary 4.502 0,0126 Control Room 0.527 0.0132 Per SRP, Section 15.7.4, Rev.1, the radiological consequences of an FHA must be within the limits of 75 rem thyroid and 6 rem whole body. The 10CFR100 limits are 300 rem thyroid and 25 rem whole body.
Additionally,10CFR50 Appendix A, General Design Criterion (GDC) 19, specifies that adequate radiation protection is to be provided to permit access and occupancy of the control room under accident conditions without personnel exposures in excess of 5 rem whole body or its equivalent to any part of the body for the duration of the accident.
The above results demonstrate that the offsite and control room doses due to an FHA in the Containment Building are well within the acceptance criteria given in SRP Section 15.7.4 and GDC 19.
A review of actual operating conditions during a refueling outage reveals additional safety margin when compared to the assumptions contained in the above analysis.
In mode 6, accidents do not result in containment pressurization, therefore, only containment closure is required. This fact, combined with containment pressure being normally less than or equal to ambient pressure during refueling outages, results in conditions that are favorable toward minimizing offsite dose consequences.
To further put into perspective the effect of having the equipment hatch open during the onset of any fuel handling accident, consider the fact that its square footage is only 0.146% of the entire containment vessel. This is in stark contrast to the assumption used in this analysis that, for all practical purposes ignores the existence of the containment building (the assumption is that all airborne activity reaching the containment is released instantaneously to the outside atmosphere).
Off Normal procedures are documented for High Airborne Activity in containment (0P-901-403) and for a fuel Handling Incident (OP-901-405). Various radiation monitors are installed to monitor containment atmosphere to provide prompt indication of high airborne activity in containment or indication of a Fuel Handling Incident.
Table A lists the monitors which monitor containment atmosphere with their respective location, type, and ranges.
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- As a compensatory measure, Waterford 3 will install during each outage a portable airborne radiation monitor at the equipment hatch which would monitor.
noble gas, particulate, and iodine. A flow direction device will also be installed to determine air flow out of the containment. The PIG which is expected to be used will have the following ranges:
Particulate (p & y) 10 10-5 pCi/cc Iodine (grossy) 10 10-3 Ci/cc Gas (gross p) 10 10-1 Ci/cc Waterford 3 will review and establish or verify the adequacy of existing procedures for the installation and operation of the portable airborne radiation monitor and for the expeditious closure of the equipment hatch.
i The proposed chango will realize safety and economic' benefits.
The intent of the current TS is to limit the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment. However, the current assumptions do not consider the large amount of people working in containment during outages. Should an FRA occur evacuation of containment will be a necessity that would be hampered by the current TS requirements. Other actions including invoking 10CFR50.54(x) could be considered to evacuate personnel.
The time required to evacuate personnel from the containment in the event of a fuel handling accident is mostly a function of the type of critical jobs ongoing and the number of personnel inside containment. Critical jobs would be immediately terminated at a point in time required to assure plant and personnel safety, and personnel searches would be conducted as required to assure all personnel are accounted for and evacuated. A conservative estimate for evacuating all personnel from containment is 45 minutes. All staging work required for closure of the equipment hatch, such as a crane, lifting rigs, and key personnel would commence immediately.
Personnel would be evacuated through both the PAL doors and the equipment hatch. The equipment hatch would be closed at that point in time at which closure could physically commence.
Any personnel remaining in containment would be evacuated through the PAL doors. A conservative estimate for closing the equipment hatch under this scenario is 60 minutes. Thus, this plan would reduce personnel dose in the event of a fuel handling accident while maintaining offsite radiation exposure within regulatory exposure limits.
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Leaving the PAL doors open will significantly reduce the wear on the doors.
Experience has shown that very frequent use of the PAL doors has resulted in accelerated wear of the PAL door components, such as door hinge pins, door seals, and the packing of equalizing valves. Thus, leaving the PAL doors open should increase the reliability of the PAL doors. Allowing the equipment and PAL doors to be open will provide for greater efficiencies in the movement of personnel and equipment without impact on the refuel critical path.
Permitting penetrations to be open for the performance of LLRT or system drain downs will further enhance outage performance by allowing activities to be conducted in parallel.
CONCLUSION A review of commercial reactor incidents involving fuel handling accidents from 1980 to the present indicates that only a few accidents involving spent fuel dropped into the core have occurred. However, none of these accidents resulted in measurable releases of activity.
Based on this review of experiences with plant operation, it appears that some elements of TS 3.9.4 and its associated basis are unnecessarily restrictive.
The overly conservative assumptions used in the analysis have a compounding effect on the dose consequences that are derived. Despite this fact, the results are still within the limits of SRP, Section 15.7.4, Rev. 1, of 75 rem thyroid and 6 rem whole body The 10CFR100 limits are 300 rem thyroid and 25 rem whole body.
Probability appears to be the reason that a fractional value of 10CFR100 is used as the acceptance limit for the offsite dose projections associated with an FHA. Although the probability of an FHA may be a frequent enough occurrence to warrant additional conservatism, the FHA analysis with its overly conservative assumption of instantaneous release of activity to the outside atmosphere, would appear to meet the threshold of a non-credible accident.
The proposed change to TS 3.9.4. contains provisions that allow the specified penetrations to be open during core alterations or movement of irradiated fuel within the containment.
The proposed provisions are considered justified when consideration is given to the compounding conservative assumptions used in the FHA analysis.
In addition, these provisions also include restrictions thM ensure timely containment closure such that acceptance criteria is met.
Administrative controls will ensure that the containment is capable of immediate closure. Additionally, trained individuals will be designated to close the containment if required. The containment will be closed as soon as personnel inside containment are evacuated, thus minimizing the release of radioactive material to the outside environment.
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Radiation Monitors Which Monitor Containment Atmosphere Nonitor Location Iran Range ARMIRE5013 Refuel MachineGM-Tube 10 10+4 mR/hr (non-safety)
ARMIRE5014 SW Stairs GM-Tube 10 10+4 mR/hr (non-safety) l
'ARMIRE5015 NE Stairs GM-Tube 10 10+4 mR/hr (non-safety)
ARMIRE5024
+46 RCB lon Chamber 20 - 5.00 E+5 mR/hr (safety)
ARMIRE5025
+46 RCB Ion Chamber 20 - 5.00 E+5 mR/hr (safety)
ARMIRE5026
+21 RCB Ion Chamber 20 - 5.00 E+5 mR/hr (safety)-
ARMIRE5027
+21 RCB Ion Chamber 20 - 5.00 E+5 mR/hr (safety) s ARMIRE5400 A
+96 RCB Ion Chamber 10+0 _ 1o+8 R/hr (safety)
ARMIRE5400 A
+96 RCB Ion Chamber 10+0 - 10+8 R/hr (safety)
PRMIRE0100 Y RAB -4 PIG P
10 10-5 Ci/cc-(safety)
I 10 10+3 pCi/cc (safety)
G 10-7 _ 10-1 Ci/cc (safety) i PRMIRE0100.1 RAB +46 PIG P
10 10-5 pCi/cc (safety)
I 10 10-3 pCi/cc (safety)
G 10-7 _ 10-1 pCi/cc (safety)
PRMIRE0100.2 RAB +46 PIG P
10 10-5 C1/cc (safety)
I 10 10-3 Ci/cc (safety)
G 10-7 _ 10-1 pCi/cc (safety)
PRMIRE0110 RAB +46 WRGM 10-7 _ 1o+5 Ci/cc (Reg. Guide 1.97) f t
e Safety Analysis The proposed change described above shall be deemed to involve a significant hazards consideration if there is a positive finding in any of the following areas:
1.
Will operation of the facility in accordance with this proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response
No The containment penetrations are not initiators of any accident. The condition of the PAL door, the equipment door and containment penetrations being open or closed has no effect on the probability of occurrence of any accident previously evaluated.
The proposed change alters assumptions made in evaluating the radiological consequences of a fuel handling accident inside the Reactor Containment Building. Allowing the containment personnel airlock and equipment hatch doors to remain open during fuel movement and core alterations does increase, however not significantly, the consequences of a fuel handling accident inside containment.
Previously, the fuel handling accident inside containment was bounded by the fuel handling accident analysis in the spent fuel pool area of the Fuel Handling Building.
The offsite and control room radiological doses due to a worst case fuel handling accident within the Reactor Containment Building, based on conservative assumptions (e.g., instantaneous release of activities from the containment building to the outside environment), have been evaluated. The evaluation has demonstrated that the offsite doses are within the limits of SRP, Section 15.7.4, Rev.1, of 75 rem thyroid and 6 rem whole body. The 10CFR100 limits are 300 rem thyroid and 25 rem whole body. The calculated control room operator doses are well within the acceptance criteria given in GDC 19.
The proposed change contains restrictions to ensure that the containment will perform its safety related function of containing fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10CFR100. Therefore, the proposed change will not involve a significant increase in the probability or consequences of an accident previously evaluated.
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2.
'Will operation of the facility in accordance with this proposed change create the possibility of a new or different type of accident from any accident previously evaluated?
Response
No.
The proposed change involves a change to the Technical Specifications that would allow the PAL door, the equipment door and certain containment penetrations to be open during fuel movement. The containment penetrations are not initiators to any analyzed or new event.
Provisions to ensure the capability to close the containment have been made in the event of a FHA.
The proposed change will not alter the operation of the plant or the manner in which the plant is operated. Therefore, the proposed change does not create the possibility of a new or different type of accident from any accident previously evaluated.
3.
Will operation of the facility in accordance with this proposed change involve a significant reduction in a margin of safety?
Response
No The offsite and control room doses due to an FHA, based on conservative assumptions, have been evaluated. The analysis demonstrated that the resultant doses are well within the appropriate acceptance limits.
Therefore, the margin of safety as defined by 10CFR100 has not been reduced.
Safety and Sianificant Hazards Determination Based on the above safety analysis, it is concluded that:
(1) the proposed change does not constitute a significant hazards consideration as defin;d by 10CFR50.92; and (2) there is a reasonable assurance that the health and safety of the public will not be endangered by the proposed change; and (3) this action will not result in a condition which significantly alters the impact of the station on the environment as described in the NRC final environmental statement.
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f NPF-38-167 ATTACHMENT A
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