ML20113B734

Application for Amend to License NPF-29,redefining Secondary Containment Boundary to Allow Encl Bldg to Be Inoperable During Upcoming Refueling Outage

ML20113B734
Person / Time
Site:	Grand Gulf
Issue date:	06/20/1996
From:	Hutchinson C ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20113B736	List: ML20113B734 ML20113B738
References
GNRO-96-00068, GNRO-96-68, NUDOCS 9606270286
Download: ML20113B734 (13)

Text

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'ENTERGY

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Port Gbson.MS 39150 Tel 601437 2800 C. R. Hutchinson June 20,1996 C7 GrW gun Ntrsar Swun U.S. Nuclear Regulatory Commission Mail Station P1-137 Washington, D.C. 20555 Attention: Document Control Desk

SUBJECT:

Grand Gulf Nuclear Station Docket No. 50-416 License No. NPF-29 Enclosure Building Isolation Requirements during Refueling Outage 8 Proposed Amendment to the Operating License GNRO-96/00068 Gentlemen:

Entergy Operations, Inc. is submitting by this letter a proposed amendment to the Grand Gulf Nuclear Station (GGNS) Operating License. The proposed change will redefine the secondary containment boundary to allow the enclosure building to be inoperable during the upcoming refueling outage. The boundary is moved by controlling the auxiliary building to enclosure building access door, ensuring that the Standby Gas Treatment (SGT) System dampers which communicate between the two buildings are closed and ensuring that SGT System will not automatically start. As a result, in the event of the design bases fuel handling accident an isolated low leakage boundary (consisting of the primary containment and the concrete auxiliary building) is automatically established.

The proposed change is very conservative with respect to the accident analysis. As discussed in Attachment 2, the fuel handling accident has been reanalyzed for the time period affected.

The reanalyzed fuel handling accidents do not credit dose mitigation by the affected engineered safety feature (ESF) systems (e.g., auxiliary building and enclosure building integrity, isolation of the containment and fuel handling area ventilation systems, and the SGT System). In addition to the assumptions of the analysis the proposed change leaves in effect a low leakage l boundary (consisting of the primary containment and the auxiliary building) which will automatically isolate in the event of the design basis fuel handling accident and requires that the SGT System be available for manual initiation when desired.

We request the Staff complete its review and approval by October 1,1996, to support the October 1996 refueling outage. This letter provides information concerning the need for the timely approval of the requested changes and requests a meeting with the Staff to discuss the resolution of any issues associated with the requested change.

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GNRO-96/00068 page 2 of 2 Based on the guidelines in 10CFR50.92, Entergy Operations has concluded that this proposed amendment involves no significant hazards considerations. Attachment 2 details the basis for this determination.

Yours truly,

. e CRH/BSF attachments: 1. Affirmation per 10CFR50.30 (1 page)

2. Proposed Change to the Operating Licensee, Enclosure Building isolation Requirements during Refueling Outage 8, (10 pages)

3. Mark-up of Affected Operating Licensee Condition (3 pages) cc: Mr. R. B. McGehee (w/a)

Mr. N. S. Reynold:: (w/a)

Mr. J. Tedrow (w/a)

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

Mr. J. W. Yelverton (w/a)

Mr. L. J. Callan (w/a)

Regional Administrator U.S. Nuclear Regulatory Cnmmission Region IV Suite 400 611 Ryan Plaza Drive Arlington, TX 76011 Mr. J. N. Donohew, Project Manager (w/2) l 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 l

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BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION l

LICENSE NO. NPF-29 DOCKET NO. 50-416 IN THE MATTER OF MISSISSIPPI POWER & LIGHT COMPANY and SYSTEM ENERGY RESOURCES, INC.

and SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION and ENTERGY OPERATIONS, INC.

AFFIRMATION 1, C. R. Hutchinson, being duly sworn, state that I am Vice President, Operations Grand Gulf

' Nuclear Station, 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; that I signed this application as the Vice President, Operations Grand Gulf Nuclear Station, of Entergy Operations, Inc.; and that the statements made and the matters set forth therein are true and correct to the best of my knowledge, information and belief.

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C. R. Hutchinson STATE OF MISSISSIPPI COUNTY OF ()/aibert?e SUBSCRIBED AND SWORN TO before mp, a Notary Public, in and fcr the County and State above named, this 20" day of A ve 1996.

(SEAL) hh/l$ntL - '"

Notary Public

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My commicslon expires: EEs@E{$753@

BONDED TilRU STEGALL fiOTMiT ET2CD I

. GNRO-96/00068 Attachment 2 page 1 of 10 l

PROPOSED CHANGE TO THE OPERATING LICENSE 1

Enck&Jre Building isolation Requirements during l Refueling Outage 8 l

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. GNRO-96/00068 4 .

Attachment 2 page 2 of 10 1

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A. AFFECTED OPERATING LICENSEE REQUIREMENTS l 1

i- The following are the Technical Specification requirements affected by the proposed change.

Limiting Condition for Operation (LCO) -

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3.3.6.2 Secondary Containment isolation Instrumentation 3.6.4.1 Secondary Containment l 3.6.4.3 Standby Gas Treatment (SGT) System t

Facility Operating License condition 2.C.(2) is being modified to identify the proposed j j changes and the associated restrictions. )

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! B. ENCLOSURE BUILDING CONDITION i

i Over the years as leaks have developed in the enclosure building roof iFe affected i

areas have been patched. As the roof has aged, the frequency of the lea 5 and thus j the required repairs has increased. Also, Grand Gulf Nuclear Station (GGNS) has j experienced severe weather which included significant hail. These two factors have i

resulted in multiple leaks through the roofing. To date, the leakage has not adversely affected the function of any safety equipment within the enclosure building nor has the leakage adversely affected the ability of the enclosure building to perform its safety j function. However, it is prudent to replace the portion of the enclosure building roofing j material that overlays the secondary containment boundary.

i C. DESIGN DESCRIPTION i

The secondary containment at GGNS consists of the auxiliary building and the j enclosure building. The auxiliary building is a reinforced concrete structure which j completely surrounds the lower portion of the primary containment and the enclosure i building is a metal-siding structure which completely surrounds the primary containment above the auxiliary building roof line.

[ The reactor vessel and the upper containment fuel racks are located in the primary j containment. While the spent fuel poolis located in the auxiliary building. Fuel 4

assemblies are transported between the spent fuel pool and the upper containment pool 3

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by use of the horizontal fuel transfer tube. The horizontal fuel transfer tube runs from the spent fuel pool located in the auxiliary building to the upper containment pool in the primary containment without entering the enclosure building.

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The primary containment equipment hatch and primary containment airlocks permit

access between the primary containment and auxiliary building. During a refueling outage the primary containment is essentially part of the auxiliary building since the primary containment equipment hatch and the primary containment personnel airlocks

are open. For the fuel handling accident in the primary containment, leakage is from the i primary containment to the auxiliary building. The enclosure building merely functions

! as a mixing volume for the source term released into the auxiliary building in the event i of a fuel handling accident in the auxiliary building or in the primary containment.

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, GNRO-96/00068

. Attachment 2 page 3 of 10 The fuel handling area and the auxiliary building ventilation systems maintain the auxiliary building at a slightly negative pressure during normal operation. These non-safety systems assure that no ambient air escapes from the fuel-handling area during fuel handling operations without first being monitored and, if needed, treated for airborne radioactivity. Upon detection of high radioactivity, the standby gas treatment (SGT) system is initiated and these systems are isolated.

The primary containment ventilation exhaust is a non-safety system which functions to discharge filtered air from the primary containment during normal shutdown operation.

Upon detection of high radioactivity this system isolates. When isolated this system can continue to be used to remove radioactive materials from the primary containment atmosphere in its recirculation mode of operation.

The SGT system maintains the secondary containment at a negative pressure and provides cleanup of the potentially contaminated secondary containment volume following a design basis accident. Following actuation, the system draws air from the auxiliary building, mixes this air with air drawn from the enclosure building, and returns I the mixed air to the enclosure building. A portion of the mixed air is exhausted via a '

charcoal filter assembly to maintain the secondary containment boundary region at a negative pressure.

D. CURRENT LICENSE BASIS

1. Fuel Handling Accident in the Auxiliary Building The fuel handling accident in the auxiliary building is evaluated in the GGNS Updated Final Safety Analysis Report (UFSAR) Section 15.7.4. This accident encompasses a fuel handling accident in the spent fuel pool. The design basis analysis is based on the Standard Review Plan (SRP) 15.7.4 and Regulatory ,

Guide 1.25. The limiting event is the drop of a channeled irradiated 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 systems and initiate the SGT subsystems. The isolation and subsequent filtration of the release limits the I transport of fission products to the environment. The specific radiological effects at the Site Boundary are identified in UFSAR Table 15.7-10.

2. Fuel handling accident in the primary containment The fuel handling accident in the primary containment is evaluated in the UFSAR Section 15.7.6. The design basis analysis is also based on the SRP 15.7.4 and Regulatory Guide 1.25. The limiting event is the drop of an irradiated fuel assembly onto irradiated fuel assemblies in the upper containment racks with the containment equipment hatch open to the auxiliary building. 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 primary containment to the environment:

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' GNRo-96/00068 Attachment 2 page 4 of 10 A) The activity released from the fuel is conservatively assumed to be completely pulled through the open equipment hatch into the auxiliary building; or B) The activity released from the fuel is largely pulled through the equipment hatch into the auxiliary building with a small amount pulled through the containment ventilation system to the environment.

The radiological effects for a fuel handling accident inside primary containment are identified in UFSAR Table 15.7-14.

For the fuel handling accidents described above, secondary containment (i.e., auxiliary building and enclosure building) working in conjunction with the SGT System limit the transport o' fission products to the environment and the associated radiological consequences (per SRP 15.7.4 guidelines) to well within the 10CFR100.11 limits. The SRP further defines the fuel handling accident limits as 75 rem thyroid and 6 rem whole body.

Loads in excess of 1140 pounds (heavy loads) are prohibited from traveling over spent fuel assemblies in the spent fuel or upper containment fuel storage pool racks. Without i appropriate controls, loads s; 1140 pounds (light 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 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.

E. REANALYSIS OF FUEL HANDLING ACCIDENT Entergy Operations, Inc. reanalyzed the fuel handling accidents for GGNS not crediting the active engineered safety feature (ESF) systems (e.g., auxiliary building and enclosure building integrity, isolation of the containment and fuel handling area ventilation systems, and the SGT System) that are currently credited in the UFSAR analyses to reduce the consequences of the analyzed events.

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. 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. The results of these analysis were submitted for NRC review for the related Technical Specification change to the fuel handling accident operational conditions dated November 9,1994. We have been informed by the Staff that independent Staff calculations reached similar conclusions as our analysis.

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. GNRO-96/00068 Attachment 2 page 5 of 10 Consistent with the analysis presented in UFSAR Sections 15.7.4 and 15.7.6, key j assumptions used in the analysis are as follows:

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1. Regulatory Guide 1.25 assumptions are followed with the exception that the ICRP 30 dose conversion factors are used for thyroid dose and whole body dose l calculations. The Regulatory Guide 1.25 assumptions include the following: j a) All of the gap fission product inventory is released after a cladding failure.

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

10% of the noble gases (excluding Kr-85) 30% of the Kr-85 inventory l

10% of the lodine inventory b) A radial peaking factor of at least 1.5 is assumed.

c) The activity released to the containment / auxiliary building is based on an overall decontamination factor of 100 for lodine for the first 23 feet of L water coverage (additional decontamination for water depths of 46 feet or i

greater is credited) and a decontamination factor of one for the noble gases (no noble gases are retained in the pool),

d) The fission products released to the containment / auxiliary building escape to the environment within two hours.

2. As discussed in UFSAR 15.7.4 and UFSAR 15.7.6, all of the fuel rods in the i

dropped bundle are assumed to fail (76 rods) with an additional 32 fuel rods failing in the impacted bundles for drops over the storage racks and 84 fuel rods for drops over.the reactor core. This large number of fuel rod failures are the result of conservative assumptions including the following: l a) It is assumed that the 3" and 4" sections of the NF500 telescoping mast and the handling tool fall from the top of the water level. This represents

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

l b) No credit is taken for drag force on the dropped object though credit is taken for the buoyancy force on the dropped object.

c) All fuel rods of a dropped bundle are assumed to fait due to bending (i.e.,

no credit is taken for lateral support provided by a fuel channel).

L d) All of the impact energy is dissipated by failing fuel rods. In other words, l each failed fuel rod only absorbs the minimum amount of energy to cause it to fail, thereby, maximizing the number of failures.

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

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GNRO-96/00068

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Anschment 2 page 6 of 10

4. Atmospheric dispersion factors (X/Q), developed using approved methodologies from Standard Review Plan Section 2.3.4, Rev.1, are used in the radiological assessment.

5. Although the containment ventilation system has non-safety related charcoal filters, no credit is taken for iodine removal.

The limiting drop without active ESF accident mitigating functions is within the dose limitations for decay periods of 12 days or more.

F. PROPOSED CHANGES This proposed amendment to the GGNS Facility Operating License will redefine the secondary containment boundary for the upcoming refueling outage during CORE ALTERATIONS and movement of non-recently irradiated fuel (i.e., fuel that has not occupied part of a critical reactor core for 12 days). This change will allow the enclosure building to be inoperable and SGT in manual mode, provided the following conditions exist:

a. All dampers communicating between the auxiliary building and the enclosure building are closed.

b. The access door between the auxiliary building and the enclosure building is closed, except when the access opening is being used for entry and exit.

c. SGT is blocked from automatic initiation.

d. SGT is available for manual initiation or the ACTIONS of LCO 3.6.4.3 are complied with.

This Facility Operating License change is an exception to the requirements of LCO 3.6.4.1, " Secondary Containment," for the enclosure building to be closed (Surveillance -

Requirement (SR) 3.6.4.1.1 and SR 3.6.4.1.2) and to have the ability to achieve and maintain the required vacuum (SR 3.6.4.1.3 and SR 3.6.4.1.4). This Facility Operating License change is also an exception to the requirements of LCO 3.3.6.2, " Secondary Containment Isolation Instrumentation," and LCO 3.6.2.3, " Standby Gas Treatment (SGT) System," for the automatic start requirements for the SGT System (SR 3.3.6.2.4, i SR 3.3.6.2.6 and SR 3.6.4.3.3).

1 G. REASON THE CHANGE IS REQUESTED FOR THE UPCOMING REFUELING OUTAGE (RFO 8)

The enclosure building has a metal decking roof which by design is sealed sufficiently to support the inleakage requirements of the secondary containment. To protect the metal decking and associated sealant (e.g., caulking) the roof decking was overlaid with approximately 2 inches of insulation, several layers of fiberglass felt, gravel, and asphalt. As previously discussed the enclosure building roof is leaking and although the leakage has not adversely affected the function of any safety equipment within the enclosure building or adversely affected the ability of the enclosure building to perform its safety function it is desired to repair the leaking roof.

. GNRO-96/00068 Attachment 2 )

page 7 of 10 ,

Replacing the enclosure building roofing material is the best option for stopping the current leaks and thereby preclude future degradation of the secondary containment boundary. Replacing the roofing material also allows for the identification and repair of any degradation of the secondary containment boundary. Because the enclosure building metal decking and associated sealant is by design sufficient to support the leaktightness requirements of the secondary containment, this work could be performed during operation or shutdown conditions. Although unlikely, the possibility exists that removing the roofing material may result in an unacceptable increase in air leakage and consequent inoperability of the enclosure building. Additionally, the metal decking could be damaged by the repair activities.

Removal and replacement of roofing material with considerations for weather uncertainties will require a period of time comparable to a significant fraction of the upcoming refueling outage schedule. Under the current requirements secondary containment is required to be OPERABLE almost the entire outage. Given the potential for creating a condition where secondary containment is inoperable, it is preferred to schedule the work with the highest risk of causing secondary containment inoperability when the enclosure building is not required to be OPERABLE. Otherwise, the enclosure building roofing repairs may lead to a significant outage extension.

H. JUSTIFICATION

SUMMARY

Thc proposed revision to the Operating License provides a mechanism to allow the en ,losure building to be inoperable and potentially undesirable SGT operation be prevented. The proposed change would only allow handling of irradiated fuel with the enclosure building inoperable following 12 days of decay. The 12 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 SRP 15.7.4 and General Design Criteria 19 without crediting the systems affected by the proposed change.

The proposed change mitigates potential effects of the enclosure building inoperability by essentially moving the secondary containment boundary. The boundary is moved by controlling the auxiliary building to enclosure building access door and ensuring that the SGT System dampers which communicate between the two buildings are closed and that SGT will not automatically start. In practice achieving and maintaining this configuration only entails controlling the enclosure building to auxiliary building door and preventing the automatic start of the SGT System because the SGT System suction dampers are normally closed and open on the start signal for the associated SGT subsystem. The effect of this change is that in the event of the design bases fuel handling accident an isolated low leakage boundary (consisting of the primary containment and the concrete auxiliary building) is established and any known leakage paths that could cause the enclosure building to leak excessively are closed prior to SGT initiation. There is no specific time frame within which the SGT System is required to be placed in service but the system must not be inoperable for any reason other than implementation of the discussed enclosure building to auxiliary building isolation unless the Technical Specification requirements for inoperable SGT subsystem (s) are met (i.e.,

up to and including halting irradiated fuel movement). These configuration changes

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. GNRO-96/00068

. . Attachment 2 page 8 of 10 prevent the potential for undesirable operation of the SGT which results in the increased release of unfiltered contaminated air into the environment.

i The proposed change is very conservative with respect to the accident analysis. As i discussed above, the reanalyzed fuel handling accident for the time period affected does not credit dose mitigation by the active engineered safety feature (ESF) systems i . (e.g., auxiliary building and enclosure building integrity, isolation of the containment and

. fuel handling area ventilation systems, and the SGT System). In excess of the i assumptions of the analyses the proposed change does leave in effect a low leakage

! boundary (consisting of the primary containment and the auxiliary building) which will j automatically isolate in the event of the design basis fuel handling accident and requires

that the SGT System be available for manual initiation when desired.

! As described in the UFSAR Chapter 15, the accidents postulated to occur during core alterations in addition to fuel handling accidents are: inadvertent criticality due to a control rod removal error or continuous control rod withdrawal error during refueling and the inadvertent loading of a fuel assembly in an improper location. These events are not 1 postulated to result in fuel cladding integrity damage. Since the only accident postulated to occur during CORE ALTERATIONS that results in a significant radioactive release is  !

the fuel handling accident, the proposed change in requirements during CORE ALTERATIONS is justified.

This proposed change does not affect the OPERABILITY requirements for any equipment during any time period during which credit need be taken for the functioning of the equipment. The proposed change also does not effect the OPERABILITY  !

requirements associated with any Operations with Potential for Draining the reactor i vessel.

The justification for this change is consistent with bases for the NRC granting a previous temporary change to the Technical Specifications for GGNS. The previous change was granted by the NRC in Amendment 22 to the Operating License, dated October 22, 1986. This change temporarily relaxed secondary containment requirements during )

control rod movements on the basis of an analysis of the effects of the event without l filtration in the SGT System. The analysis showed that the control room and offsite doses were within the acceptance criteria in the SRP,10CFR50 Appendix A GDC 19 and 10 CFR Part 100 and were, as a result, acceptable.

l. NO SIGNIFICANT HAZARDS CONSIDERATIONS i l

This proposed amendment to the Grand Gulf Nuclear Station (GGNS) Technical i Specifications (TS) revises the operability requirements for the enclosure building and l the Standby Gas Treatment (SGT) System during CORE ALTERATIONS and the ' j handling of non-recently irradiated fuel (i.e., fuel that has not occupied part of a critical  !

reactor core for 12 days) during the upcoming refueling outage.  !

l The Commission has provided standards for determining whether a no significant l hazards consideration exists as stated in 10CFR50.92(c). A proposed amendment to an a 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)

. GNRO-96/00068 Attachment 2 page 9 of 10 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 ameridment 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.

The equipment affected by the proposed change is not considered an initiator to any previously analyzed accident, therefore, inoperability of the equipment does not increase the probability of any previously evaluated accident. '

As described in Updated Final Safety Analysis Report Chapter 15, the accidents postulated to occur during core alterations in adclition to fuel handling accidents are:

inadvertent criticality due to a control rod removal error or continuous control rod withdrawal error during refueling and the inadvertent loading of a fuel assembly in an improper location. These events are not postulated to result in fuel cladding integrity damage. The only accident postulated to occur during CORE ALTERATIONS that results in a significant radioactive release is the fuel handling accident. The proposed requirements 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 contcquences 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 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.

The leaktightness of the enclosure building does not affect the function of any plant system other than the ability of the SGT System to ensure the secondary containment is at the specified pressure. The proposed change in normal SGT System alignment by defeating the automatic start feature of the SGT System and the inability to ensure secondary containment is at the specified pressure does not affect the operation of any plant system or component. The SGT System is not relied upon to provide normal or accident cooling to plant systems or components.

The function of the enclosure building and the SGT System is only to mitigate the release of radioactivity to the environment in the event of an accident.

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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GNRO-96/00068

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, page 10 of 10 l

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

The proposed changes continue to ensure that the radiological consequences are at or below the current GGNS licensing limit. Safety margins and analytical conservatisms have been evaluated and are well understood. Substantial margins are retained to ensure that the analysis adequately bounds all postulated event scenarios. The current margin of safety is retained.

Specifically, the margin of safety for the fuel handling accident is the difference between the 10CFR100 limits and the licensing limit defined by NUREG 0800, Section 15.7.4. With respect to the control room personnel doses, the margin of safety is the difference between the 10CFR100 limits and the licensing limit defined by 10CFR50, Appendix A, Criterion 19 (GDC 19). The proposed applicability continues to ensure that the whole-body and thyroid doses at the exclusion area and low population zone boundaries as well as control room, doses are at or below the corresponding licensing limit. The margin of safety is unchanged; therefore, the proposed changes do not involve a significant reduction in a margin of safety.

In excess to the margin of safety supplied by the licensing limits of NUREG 0800 and GDC 19, the proposed change incorporates an additional layer of conservative requirements. The proposed change leaves in effect a redefined secondary containment boundary which will provide a low leakage boundary (consisting of the primary containment and the auxiliary building) by automatically isolating in the event of the design basis fuel handling accident and requires that the SGT System be available for manualinitiation when desired. These requirements will ensure that doses will be sven lower than those calculated.

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

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

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