ML20079B443

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Application for Amend to License NPF-29,revising TS Associated W/Various ESF Sys Following Design Basis Fuel Handling accident.Mark-up of Current Ts,Proposed TS & mark- Up Improved TS Encl
ML20079B443
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 11/09/1994
From: Hutchinson C
ENTERGY OPERATIONS, INC.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20079B445 List:
References
GNRO-94-00131, GNRO-94-131, NUDOCS 9501060106
Download: ML20079B443 (72)
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Port Gibson. MS 39150 Td 601437 2800

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C. R. Hutchinson Vce PmMW Ga Gu Mik4tf $1!I1K#1 November 9,1994 l

U.S. Nuclear Regulatory Commission i

Mail Station P1-37 Washington, D.C. 20555 Attention: Document Control Desk

SUBJECT:

Grand Gulf Nuclear Station Unit 1 Docket No. 50-416 License No. NPF-29 Fuel Handling Accident Operational Conditions Proposed Amendment to the Operating Licensa (PCOL-93/08)

GNRO-94/00131 Gentlemen:

Entergy Operations, Inc. is submitting by this letter a proposed amendment to the Grand Gulf Nuclear Station (GGNS) Operating License. The proposal revises those specifications associated with various engineered safety feature (ESF) systems following a design basis fuel handling accident. The proposed changes affect conditions where irradiated fuelis handled in the primary or secondary containment and when fuel is handled over the reactor vessel with fuelin the vessel. These changes are based on the recent re-analysis of the fuel handling accident for GGNS. Specifically, the proposed changes add a new definition for RECENTLY 1RRADIATED fuel, and revises ACTIONS for Containment isolation Instrumentation, Radiation Monitoring Instrumentation, Secondary Containment Integrity, Secondary Containment Automatic Isolation Dampers / Valves, Standby Gas Treatment System, Standby Service Water, Ultimate Heat Sink, Control Room Emergency Filtration System, AC and DC Electrical Power Systems - Shutdown, and Electrical Power Distribution Systems - Shutdown.

This proposed amendment is being submitted as part of the cost beneficiallicensing action (CBLA) program established within NRR where increased priority is granted to licensee requests for changes requiring staff review that involve high cost without a commensurate safety benefit. Entergy developed the proposed changes to decrease the operational burden placed on outage resources. We expect accrued cost reductions in excess of $500,000 over the remaining operating life of the plant without reducing safety margin. Entergy requests that priority for this CBLA item be consistent with receiving NRC approval concurrent with approval of the GGNS Improved Technical Specifications (ITS) during the first quarter of 1995 in order to facilitate planning for our seventh refueling outage. However, for review and comparison we have included discussion of changes to the current GGNS Technical Specifications (CTS) and mark-ups of both the CTS and ITS.

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GNRO-94/00131 Page 2 of 3 provides a detailed description of the pmposed changes, justification, and the No Significant Hazards Considerations. Attachment 3 is a copy of the marked-up CTS pages, is an information copy of the retyped CTS pages and Attachment 5 is a copy of the marked-up ITS.

In accordance with the provisions of 10CFR50.4, the signed original of the requested amendment is enclosed. This amendment request has been reviewed and accepted by the Plant Safety Review Committee and the Safety Review Committee.

Based on the guidelines in 10CFR50.92, Entergy Operations has concluded that this propor ad amendment involves no significant hazards cons.iderations. Attachment 2 details the basis,or this determination.

Yo s truly, H/ML )/am i

ttac ents:

Affirmation per 10CFR50.30

2. GGNS PCOL-93/08
3. Mark-up of Affected Current Technical Specification Pages
4. Proposed Technical Specifications Pages - Information Only
5. Mark-up of improved Technical Specification Pages cc:

Mr. J. E. Tedrow (w/a)

Mr. H. W. Keiser (w/a)

Mr. R. B. McGehee (w/a)

Mr. N. S. Reynolds (w/a)

Mr. H. L. Thomas (w/o)

Mr. Stewart D. Ebneter (w/a)

Regional Administrator U.S. Nuclear Regulatory Commission Region Il 101 Marietta St., N.W., Suite 2900 Atlanta, Georgia 30323 Mr. P. W. O'Connor, Project Manager (w/2)

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 13H3 Washington, D.C. 20555 Dr. Eddie F. Thompson (w/a)

State Health Officer State Board of Health P.O. Box 1700 Jackson, Mississippi 39205

i Attachm:nt 1 to GNRO-94/00131 Fage 1 of 1 BEFORE THE -

. UNITED STATES NUCLEAR REGULATORY COMMISSION t

~ LICENSE NO. NPF-29 l

DOCKET NO. 50-416 L

i IN THE MATTER OF MISSISSIPPI POWER & LIGHT COMPANY and SYSTEM ENERGY RESOURCES, INC.

I and SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION and l

ENTERGY OPERATIONS, INC.

AFFIRMATION I, M. J. Meisner, state that I am Director, Nuclear Safety & Regulatory Affairs, GGNS of Entergy Operations, Inc.; that on behalf of Entergy Operations, Inc., System Energy Resources, Inc., and South Mississippi Electric Power Association I am authorized by Entergy

[

Operations, Inc. to sign and file with the Nuclear Regulatory Commission, this application for amendment of the Operating License of the Grand Gulf Nuclear Station; that I signed this.

application as Director, Nuclear Safety & Regulatory ffairs, GGNS of Entergy Operations, i

inc.; and that the statements made and the matter set forth ierein are true and correct to the i

best of my knowledge, informatian and belief.

-M

. M. J.

ei er STATE OF MISSISSIPPI COUNTY OF CLAlOOi;NE Warre iM 94p SUBSCRIBED AND S N TO befojr me, a Notary Public, in and for the County and State i

day of Yahn (N' A >

.1994.

abovo named, this L

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Notary Public My commission expires:

E~S$[NoSEEINNIEIN7 matwn Tumu sist.Au NOTAH SElmCE

, to GNRO 94/00131 Page 1 of 11 PROPOSED CHANGE TO THE OPERATING LICENSE FUEL HANDLING ACCIDENT OPERATIONAL CONDITIONS (GGNS PCOL 93/08)

Gr:nd Gulf Nucle:r Station Attachm:nt 2 to GNRO-94/00131 PCOL 93/08 Page 2 of 11 A.

SUBJECT:

Fuel Handling Accident Operational Conditions Technical Specifications: 1.35a (new), Table 3.3.2-1, Table 4.3.2.1-1, Table 3.3.7.1-1, 4.3.7.1-1, 3.6.6.1, 3.6.6.2, 3.6.6.3, 3.7.1.1, 3.7.1.3, 3.7.2, 3.8.1.2, 3.8.2.2, 3.8.3.2 and bases for TS 3/4.9.4.

Affected Pages:

1-7, section 3/4 pages 3-14,3-26,3-59*,3-60*,3-61*,3-62*,6-48, 6-49, 6-55, 7-1, 7-4, 7-5*, 8-9, 8-14, 8-17, 8-18, and B 3/4 9-1.

  • Indicated pages also show related changes previously submitted with the improved Tech Specs via PCOL 93/11 R2.

B. DISCUSSION:

Following reactor shutdown, decay of the short-lived fission products greatly reduces the fission product inventory present in irradiated fuel. Technical Specification 3.9.4 requires a l

24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period of reactor suberiticality prior to fuel movement. This delay is an assumption I

in the fuel handling accident analysis. The proposed changes are based on a longer decay period (12 days) and take advantage of the reduced radionuclide inventory available for release in the event of a fuel handling accident. The proposed changes redefine the operability requirements for selected engineered safety feature (ESF) systems such that these systems are only required to be operable during this 12 day decay period. The affected systems are secondary contair ment and the associated isolation and radiation monitoring instrumentation, secondary containment isolation damper / valves, the standby gas treatment system (SGTS), service water systems, control room emergency filtration system, AC and DC Electrical Power Systems - Shutdown, and Electrical Power Distribution Systems - Shutdown.

Implementation of the proposed changes will have a significant impact on outage activities at GGNS resulting in reduced outage costs and increased flexibility with no impact on safety margin. Currently, moving large equipment into secondary containment such as chemical-decon equipment or safety-relief valves must either be delayed or moved through an alternate entrance. Because of the high level of modification work, maintenance, and repair activities during outages, wear and tear on access doors to secondary containment causes the doors to frequently break down which creates a bottle-neck situation for processing personnel and equipment in and out of the radiological control area. When this occurs, contract personnel must be rerouted through other, less convenient access paths.

Also, additional door guards are typically employed during outages to ensure inat incidents of doors being left open are minimized. These factors coupled with the increased flexibility for scheduling testing and maintenance activities on secondary containment valves, dampers, and instrumentation can result in accrued cost reductions in excess of $500,000 over the remaining operating life of the plant and allow outage resources to be directed elsewhere.

Original License Basis The fuel handling accident in the auxiliary building is evaluated in the GGNS UFSAR Section 15.7.4. The design basis analysis is based on the Standard Review Plan 15.7.4 and Regulatory Guide (RG) 1.25. The limiting event is the drop of a channeled irradiated

Grand Gulf Nucle:r Station Attrchm::nt 2 to GNRO-94/00131 PCOL 93/08 Page 3 of 11 fuel assembly onto stored spent fuel bundles. The cause of this event is a failure of the fuel assembly lifting mechanism. The radioactive release causes high radiation signals to isolate the normal ventilation system and initiate the standby gas treatment system. The Technical Specifications define operability, set points, closure times, and surveillance intervals for the fuel handling area ventilation exhaust and pool sweep radiation monitors, the SGTS, and the secondary containment automatic isolation damper / valves and the associated electrical power systems. These systems limit the transport of fission products I

to the environment such that the radiological effects at the Site Boundary are approximately 1.7 rem whole body and 2.3 rem thyroid.

The fuel handling accident in the containment is evaluated in the GGNS UFSAR Section 15.7.6. The design basis analysis is also based on the Standard Review Plan (SRP) 15.7.4 and Regulatory Guide (RG) 1.25. The limiting event is the drop of an irradiated fuel assembly onto the reactor core with the containment equipment hatch open to the secondary containment. The cause of this event is a failure of the fuel assembly lifting mechanism. The UFSAR evaluates two cases for transporting the radioactivity released from containment to the environment: 1) The activity released from the fuelis conservatively assumed to be completely pulled through the open equipment hatch into the auxiliary building; and 2) The activity released from the fuel is partially pulled through the equipment hatch into the auxiliary building and partially pulled through the containment ventilation system to the environment. For case 2, the radiological effects at the Site Boundary are approximately 1.7 rem whole body and 2.5 rem thyroid. Case 1 is slightly less.

For the fuel handling accidents described above, secondary containment integrity, isolation of the containment and fuel handling area ventilation systems, working in conjunction with the SGTS limit the transport of fission products to the environment and the associated radiological consequences (per SRP 15.7.4 guidelines) to well within the 10 CFR 100.11 limits. The SRP further defines the fuel handling accident limits as 75 rem thyroid and 6 rem whole body. Because these systems are directly related to mitigating the release of radioactive material and are part of the primary success path for the design basis fuel handling accident, appropriate operating restrictions are imposed by the Technical Specifications.

Loads in excess of 1140 pounds are prohibited by the Technical Specifications from traveling over spent fuei assemblies in the spent fuel or upper containment fuel storage pool racks. Without appropriate controls, loads weighing less than 1140 pounds (light l

loads) of sufficient impact energy could result in exceeding the SRP 15.7.4 dose limitations if dropped on irradiated fuel assemblies. This issue was identified via LER-88/016-1 (AECM-89/0025) dated February 1,1989 (Final Report). The resolution to this LER established administrative controls that involve height / weight limits that control the impact energy of light loads to assure that, in the unlikely event of a drop over irradiated fuel, offsite radiological consequences would be limited to the SRP 15.7.4 limits. This proposed i

amendment applies the same criteria to establish controls on the handling of RECENTLY IRRADIATED fuel, thereby allowing irradiated fuel and " light loads" to be controlled on the same basis.

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Grand Gulf Nucle:rStati:n Attachm:nt 2 t) GNRO-94/00131 PCOL 93/08 Page 4 of 11 Reanalysis of Fuel Handling Accident Entergy Operations, Inc. recently reanalyzed the Fuel Handling Accidents for Grand Gulf Nuclear Station. The reanalysis was performed to incorporate ICRP 30 dose conversion factors, consideration of the drop of the fuel handling tool, updated atmospheric dispersion factors (x/Q factors), and the impact of not crediting various engineered safety featt.re (ESF) systems that are currently used to reduce the consequences of the analyzed events.

Implementation of the ICRP 30 dose conversion factors for the GGNS Loss of Coolant Accident (LOCA) analysis, revised x/Q factors, and other significant dose calculation methodology and assumption changes were incorporated into the GGNS license basis under 10 CFR 50.59. Entergy presented the new license basis during a meeting with the NRC Staff on April 6,1993.

The ICRP 30 conversion factors are based on additional empiricalinformation and improved understanding of radiation effects. Atmospheric dispersion values (x/Q factors) are based on meteorological conditions in the area surrounding the site. GGNS has been gathering this data since before the plant began operation and has updated the various factors which rely on that data. The major effect of revised meteorological data is on the x

/Q factors which become slightly less favorable.

The reanalysis evaluated the Fuel Handling Accident in the auxiliary building and in containment. Precursors for these events are unchanged from that described in the UFSAR; however, the analysis was expanded to evaluate the effects of various decay time periods beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in conjunction with the assumption that those systems used to mitigate the accident as described in the UFSAR are not available.

The analysis approach was to calculate the whole body and thyroid dose due to a single fuel rod failure. Using the dose due to a single fuel rod failure and the regulatory dose limits (75 rem thyroid,6 rem whole body), the maximum number of rod failures which could occur without exceeding the regulatory dose limits was calculated. Based on the calculated impact energy to the cladding which would result in a fuel rod failure, the limiting drops were evaluated at various decay periods.

The analysis demonstrated that for the worst case drop the regulatory dose limitations of SRP 15.7.4 are satisfied for decay periods of 12 days or more without credit for the ESF systems discussed above. On or before the 12th day following shutdown, the thyroid dose proved to be limiting with a postulated thyroid dose exceeding the 75 rem limit.

Key assumptions used in the analysis are as follows:

Regulatory Guide 1.25 [Ref. 6) assumptions are followed with the exception that the o

ICRP 30 dose conversion factors are used for thyroid dose and whole body dose calculations.

o in accordance with NUREG/CR-5009, a release fraction of 12% was applied to 1-131 for extended bumup fuel.

o Credit for 12 months decay time for GE fuel previously discharged.

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Gr~nd Gulf Nucle:r Station to GNRO-94/00131 PCOL 93/08 Page 5 of 11 it is conservatively assumed that the 3" and 4" sections of the NF500 telescoping o

mast and the handling tool fall from the top of the water level.

impact energy associated with a struck fuel assembly is absorbed in the entire o

volume of the fuel rod cladding. No credit is taken for non-cladding items such as tie-plates, water rods, or fuel pellets.

Credit is taken for the buoyancy force on the dropped object. No credit is taken for o

drag force on the dropped object.

All fuel rods of a dropped bundle are assumed to fait due to bending (i.e., no credit o

is taken for lateral support provided by a fuel channel).

Atmospheric dispersion factors (x/Q), developed using approved methodologies o

from Standard Review Plan Section 2.3.4, Rev.1, are used in tw radiological assessment.

Per RG 1.25, all of the gap fission product inventory is released after a cladding o

failure. This gap inventory, based on the fraction of the total fission products, is as follows:

10% of the noble gases (excluding Kr-85) 30% of the Kr-85 inventory 10% of the lodine inventory Per RG 1.25, the activity released to the containment / auxiliary building is based on o

an overall decontamination factor of 100 for lodine for each 23 feet of water coverage (e.g., for water depths of 46 feet or greater a decontamination factor of 10,000 is used) and a decontamination factor of one for tne noble gases (no noble gases are retained in the pool).

Per RG 1.25, the fission products released to the containment / auxiliary building o

escape to the environment within two hours.

The limiting drop without accident mitigating functions is within the dose limitations for decay periods of 12 days or more. Therefore, the bounding decay period of 12 days was chosen as the basis for the proposed Technical Specifications changes. Based on these results operability requirements were established for Containment isolation Instrumentation, Radiation Monitoring Instrumentation, Secondary Containment Integrity, Secondary Containment Automatic isolation DampersNalves, Standby Gas Treatment System, Standby Service Water, Ultimate Heat Sink, Control Room Emergency Filtration system, AC and DC Electrical Power Systems - Shutdown, and Electrical Power Distribution Systems - Shutdown.

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Grnnd Gu'f Nucirr Station to GNRO-94/00131 PCOL 93/08 Page 6 of 11 Proposed Changes This proposed amendment to the Grand Gulf Nuclear Station (GGNS) Technical Specifications (TS) revises those specifications associated with handling irradiated fuel in the primary or secondary containment. The purpose is to establish a point where operability of those systems typically used to mitigate the consequences of a fuel handling accident is no longer required to meet the current license basis offsite dose limitations (75 rem thyroid,6 rem whole body). Specifically, the proposal adds a new definition for irradiated fuel that contains sufficient fission products to require operability of accident mitigation systems to meet the accident analysis assumptions and revises the operability requirements for Containment isolation Instrumentation, Radiation Monitoring instrumentation, Secondary Containment integrity, Secondary Containment Automatic isolation DampersNafves, Standby Gas Treatment System, Standby Senrice Water, Ultimate Heat Sink, Control Room Emergency Filtration System, AC and DC Electrical Power Systems - Shutdown, and Electrical Power Distribution Systems - Shutdown. The proposed changes are summarized below.

1) A new definition is added as follows:

1.35a RECENTLY IRRADIATED fuel shall be any nuclear fuel assembly that has occupied part of a critical reactor core within the previous 12 days.

2) Footnote

i "When handling RECENTLY IRRADIATED fuelin the primary or secondary containment and during operations with a potential for draining the reactor vessel."

"RECENTLY IRRADIATED" replaces " irradiated" for applicability footnote

  • in TS 3.7.1.1,3.7.1.3,3.8.1.2,3.8.2.2 and 3.8.3.2. Also, ACTION statements are revised, as appropriate, to reflect the proposed applicability footnotes. Note that the markup for the proposed improved Technical Specifications (ITS) differs slightly to be consistent with the ITS terminology and NUREG 1434.
3) The bases of TS 3/4.9.4, Decay Time, is revised to read as follows:

The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> minimum requirement for reactor suberiticality prior to fuel movement ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. During the 12 day interval used in definition 1.35a for RECENTLY l

IRRADIATED fuel, selected ESF systems are required to limit the radiological consequences of a fuel handling accident to within regulatory limits. These decay times are consistent with the assumptions used in the accident analyses.

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Grand Gulf Nucirr Station to GNRO-94/00131 PCOL 93/08 Page 7 of 11 C. JUSTIFICATION:

1) The new definition for RECENTLY IRRADIATED fuel provides a mechanism for applying a cutoff in fission product decay to various specifications where the concept applies. The twelve day period has been shown by analysis to provide sufficient decay such that, assuming the design basis fuel handling accident, radiological consequences are within the acceptance criteria of NUREG 0800, Section 15.7.4 [Ref. 5] and General Design Criteria 19 [Ref. 3].
2) The revised footnotes incorporate the newly defined term to establish operational conditions where specific activities represent situations where significant radioactive releases can be postulated. During MODE 4 or 5, these are:

a) When handling RECENTLY IRRADIATED fuelin the primary or secondary containment.

b) During operations with a potential for draineg the reactor vessel.

The footnotes for service water systems and the electrical power systems only includes handling fuel because the other conditions are implicitly included by the requirement that these systems be operable in OPERATIONAL CONDITIONS 4 and 5.

The revised footnotes redefine the LCOs' applicability for instrumentation / devices that initiate alarms, isolate containment, and provide for filtration systems, including support systems, that mitigate the radiologicalimpact of fuel handling accidents. The proposed applicability is consistent with the fuel handling accident assumptions. The applicability to RECENTLY IRRADIATED fuel bounds events where this fuel is dropped onto other RECENTLY IRRADIATED fuel. As described in the UFSAR [Ref. 2], the accidents postulated to occur during core alterations are: inadvertent criticality due to a control rod removal error or continuous control rod withdrawal error during refueling and the inadvertent loading and operation of a fuel assembly in an improperlocation. These events are not postulated to result in fuel cladding integrity damage during shutdown.

Since the only accident postulated to occur during CORE ALTERATIONS that results in a significant radioactive release is the fuel handling accident, the relationship to CORE ALTERATIONS is not appropriate. Therefore, the proposed LCO applicability for handling RECENTLY IRRADIATED fuel assemblies is justified. The applicability related to operations with a potential for draining the reactor vessel is unaffected by the proposed changes.

3) The bases of TS 3/4.9.4, Decay Time, is revised to include the basis for the 12 day period used in the new definition for RECENTLY IRRADIATED fuel. The proposed wording is consistent with bases of the changes proposed in items 1-3 above. This location in the bases to describe the concept of RECENTLY IRRADIATED fuel was chosen to minimize changes to the current TS bases and due to the similarity with the TS 3/4.9.4 bases.

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Grand Gulf Nuclear Station to GNRO-94/00131 PCOL 93/08 Page 8 of 11 Supplemental Justification The proposed changes to the Technical Specifications are based, for the most part, on reanalysis of the GGNS fuel hand!!ng accident. Each of the assumptions used in the analysis as well as the methodology mis notably conservative relative to the conditions typically present when fuel handling occurs. In addition to the conservative assumptions of Regulatory Guide 1.25, additional conservatism is included in the GGNS reanalysis of this event.

For fuel assemblies struck by a dropped object, it is assumed that the impact energy is absorbed by the fuel rod cladding. No credit is taken for non-clad items such as tie plates, water rods, or fuel channels. In addition, all of the impact energy is dissipated by failing fuei rodt in other words, each failed fue; rod only absorbs the minimum amount of energy to cause it to fail, thereby, maximizing the number of failures.

The analysis assumes that the handling tool faMs from the top of the water levelin

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containment. This represents a significant fraction (approximately 40%) of the impact energy that is absorbed by the impacted fuel.

Although the containment ventilation system has charcoal filters, no credit is taken for iodine removal.

No credit is taken for irradiation strengthening of the 9x9-5 fuel cladding. The yield

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strength of the cladding increases by approximately 10% after only 10 days of irradiation. Although no specific analysis was performed, it is anticipated that, for periods shorter than 10 days, the low fission product buildup would be more than enough to compensate for the reduced clad strength.

Each of the above conservatisms has a significant impact on the radiological consequences of a fuel handling accident even considering the worst case assumptions imposed by Reg. Guide 1.25.

D. NO SIGNIFICANT HAZARDS CONSIDERATIONS:

This proposed amendment to the Grand Gulf Nuclear Station (GGNS) Technical Specifications (TS) revises those specifications associated with handling irradiated fuel in the primary or secondary containment and CORE ALTERATIONS. Specifically, the proposal adds a new definition for irradiated fuel that contains sufficient fission products to require operability of accident mitigation systems to meet the accident analysis assumptions and revises the operability requirements for Containment isolation Instrumentation, Radiation Monitoring instrumentation, Secondary Containment Integrity, Secondary Containment Automatic isolation Dampers / Valves, Standby Gas Treatment System Standby Service Water, Ultimate Heat Sink, Control Room Emergency Filtration System, AC and DC Electrical Power Systems - Shutdown, and Electrical Power Distribution Systems - Shutdown. ACTION statements are revised, as appropriate, to reflect the new applicability footnotes. In addition, administrative changes are proposed to relocate some requirements to plant procedures in accordance with the improved Standard Technical Specifications presented in NUREG 1434.

Grand Gulf Nuclear Station to GNRO-94/00131 PCOL 93/08 Page 9 of 11 The Commission has provided standards for determining whether a no significant hazards consideration exists as stated in 10CFR50.92(c). A proposed amendment to an operating license involves a no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

Entergy Operations Inc. has evaluated the no significant hazards considerations in its request for a license amendment. In accordance with 10CFR50.91(a), Entergy Operations Inc. is providing the analysis of the proposed amendment against the three standards in 10CFR50.92(c). A description of the no significant hazards considerations determination follows:

1. The proposed changes do not significantly increase the probability or consequences of an accident previously evaluated.

a) The proposed definition of RECENTLY IRRADIATED fuelis used to establish operational conditions where specific activities represent situations where significant radioactive releases can be postulated. These operational conditions are consistent with the design basis analysis. Because the equipment affected by the revised operational conditions is not considered an initiator to any previously analyzed accident, inoperability of the equipment cannot increase the probability of any previously evaluated accident. The proposed applicability in conjunction with existing administrative controls on light loads, bounds the conditions of the current design basis fuel handling accident analysis which concludes that the radiological consequences are within the acceptance criteria of NUREG 0800, Section 15.7.4 and General Design Criteria 19. Therefore, the proposed changes do not significantly increase the probability or consequences of any previously evaluated accident.

Based on the above, the proposed changes do not significantly increase the probability or consequences of any accident previously evaluated.

2. The proposed changes would not create the possibility of a new or different kind of accident from any previous analyzed.

a) The proposed definition is used to establish operational conditions where specific activities represent situations where significant radioactive releases can be postulated. These operational conditions are consistent with the design basis analysis. The proposed changes do not introduce any new modes of plant operation and do not involve physical modifications to the plant. Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previous analyzed.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously analyzed.

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~ Grand Gulf Nuclear Station to GNRO-94/00131 PCOL 93/08 Page 10 of 11 :

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3.' ' The proposed changes do not involve a significant reduction in a margin of safety.

l a) The revised definition is used to establish operational conditions where specific

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activities represent situations where significant radioactive releases can be -

j postulated. These operational coriditions are consistent with the design basis '

analysis and are established such that the radiological consequences are at or i

below the current GGNS licensing limit. Safety margins and analytical conservatisms have been evaluated and are well understood. Substantial margins i

are retained to ensure that the analysis adequately bounds all postulated event

.i scenarios. The proposed change only eliminates the excess margin from the analysis. The current margin of safety is retained.

Specifically, the margin of safety for the fuel handling accident is the difference

]

between the 10 CFR 100 limits and the licensing limit defined by NUREG 0800, j

Section 15.7.4. With respect to the control room personnel doses, the margin of safety is the difference between the 10 CFR.100 limits and the licensing limit i

defined by 10 CFR 50, Appendix A, Criterion 19 (GDC 19). Excess margin is the difference between the postulated doses and the corresponding licensing limit.

The proposed applicability continues to ensure that the whole-body and thyroid I

doses at the exclusion area and low population zone boundaries as well as control j

room are at or below the corresponding licensing limit. The margin of safety is.

l unchanged; therefore, the proposed changes do not involve a significant reduction '

i in a margin of safety.

Therefore, the proposed changes do not result in a significant reduction in a margin of I

safety.

l Based on the above evaluation, operation in accordance with the proposed amendment involves no significant hazards considerations, j

E.

REFERENCES:

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1. Grand Gulf Nuclear Station Unit 1 Technical Specifications and Bases, Updated j

through Amendment 113.

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2. Grand Gulf Nuclear Station Final Safety Analysis Report, Updated through Revision 8, l

Chapter 15.

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3. 10 CFR 50, Appendix A, General Design Criteria 19.
4. NUREG 1454, Standard Technical Specifications, General Electric BWR/6 Plants, Revision 0, September 29,1992.
5. NUREG 0800, (Standard Review Plan), Section 15.7.4, " Radiological Consequences of Fuel Handling Accidents," Revision 1, July 1981.

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Grand Gulf Nuclear Station to GNRO-94/00131

.~ PCOL 93/08 Page 11 of 11

6. Regulatory Guide 1.25, "Assumptior,s Used for Evaluating the Potential Radiological 1

Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility j

for Boiling and Pressurized Water Reactors", 3/23/72.

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9 MARKED-UP CURRENT TECHNICAL SPECIFICATIONS PAGES FUEL HANDLING ACCIDENT OPERATIONAL CONDITIONS i

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