ML20196G690
| ML20196G690 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 06/23/1999 |
| From: | Eaton W ENTERGY OPERATIONS, INC. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20196G695 | List: |
| References | |
| GNRO-99-00049, GNRO-99-49, NUDOCS 9907010247 | |
| Download: ML20196G690 (22) | |
Text
e-En r operations. Inc.
Port Grbson, MS 39150 '
Tel G01437 C409 i Fax 601437 2795 l William A. Eaton
$$a\??"'
GrandGulf NuclearStapon June 23, 1999
- U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555.
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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 License, LDC 1999-051 j
Reference:
GNRI-96/00213, Issuance of Amendment No.129 to Facility Operating License !
No. NPF Grand Gulf Nuclear Station, Unit 1 (TAC No. M95762), dated October 18,1996.
GNRO-99/00049 -
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Gentlemen:
Entergy Operations, Inc. (EOl) is submitting by this letter a proposed amendment to the Grand Gulf Nuclear Station (GGNS) Operating License. EOl has been working with the Staff, on both
- a plant-specific as well as an industry-wide generic basis, since November 1994 to resolve issues associated with; the proposed change. The remaining conc arns associated with the issue have been resolved'as demonstrated by the issuance of Amendment 102 to the Perry Nuclear Power Plant (PNPP) Operating License addressing this issue. This requested change is consistent with the changes approved for PNPP and the industry proposed change to the Technical Specification NUREGs, TSTF-51. A key difference between those two submittals and the proposed GGNS changes is that GGNS has retained the operability requirements during gj fuel movement for the control room fresh air and air conditioning systems and the electrical Qo power systems. 4 g "/ \
The requested changes are important to GGNS. They permit substantially more flexibility in the scheduling of outage activities that can translate into both greater safety and improved resource and equipment utilization. GGNS has evaluated the upcoming RF10 outage schedule and has determined that these changes could result in a net saving of one day of the overall outage duration. On this basis, and considering the recent PNPP approved submittal on this topic,
'.GGNS respectfully requests your review and approval of these changes by September 30, 1999.- This date will permit us to implement the changes and realize the full benefit during the refueling outage.
9907010247 990623 PDR ADOCK 05000416 P PDR-L
GNRO-99/00049 Page 2 of 3 !
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The requested Technical Specification changes revise those specifications associated with various engineered safety feature systems which need no longer be credited following a design basis fuel handling accident. The proposed changes affect conditions where irradiated fuel is handled in the primary or secondary containment and certain specifications related to performing core alterations. These changes are based on the revised analysis of the fuel handling accident for GGNS.
Attachment 2 provides a detailed description of the proposed changes and a discussion of the technical and risk supporting its justification. Attachment 3 is a copy of the marked-up Operating License and TS pages.
Consistent with the PNPP submittal, Attachment 2 to this letter also includes a discussion of shutdown safety controls that will be utilized during the outage. These shutdown safety controls I will be in effect pending revision to NUMARC 93-01, Section 11.2.6, " Safety Assessment for Removal of Equipment from Service During Shutdown Conditions." As discussed further in Attachment 2, the draft NUMARC 93-01 concepts will be utilized at GGNS for controlling the removal from service of systems, structures and components that are currently required by Tr;chnical Specifications during core alteration / fuel handling periods.
Based on the guidelines in 10CFR50.92, Entergy Operations has concluded that this proposed amendment involves no significant hazards considerations. Attachment 2 also provides the l basis for this determination. '
Yours truly, ,
/FGB attachments: 1. Affirmation per 10CFR50.30 '
- 2. Proposed Change to the Operating Licensee, Fuel Handling Operational Conditions
- 3. Mark-up of Affected Operating License Conditions, Technical Specifications, and Bases cc: (See Next Page)
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GNRO-99/00049 Page 3 of 3 cc: Ms. J. L. Dixon-Herrity, GGNS Senior Resident (w/a)
Mr. L. J. Smith (Wise Carter) (w/a)
Mr. N. S. Reynolds (w/a)
Mr. H. L. Thomas (w/o)
Mr. E. W. Merschoff (w/a) i Regional Administrator l
U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 ,
Mr. S. P. Sekerak, NRR/DLPM/PD IV-l (W/2)
U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 13D18 11555 Rockville Pike Rockville, MD 20852-2378 l
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- 1. .
I GNRO-99/00049 bec: - File (LCTS/RPTS) (w/a) -
File (Hard Copy) (w/a)
File (Central) (w/a) ( 5 )
ccmail: J Mr. W. B. Abraham Mr. C. D. Holifield i Mr. G. R. Ashley (ANO - LIC) ' Mr. W. K. Hughey i Mr. W. J. Beck (RB - LIC) Mr. M. G. Hurley - i Mr. S. J. Bethay (ECH - LIC) Mr. C. R. Hutchinson (ANO - VP)
Mr. J. G. Booth IEA (ANO - OE)
Mr C. A. Bottemiller
- Mr. D. E. James (ANO - LIC)
Mr. P. W. Brewer Mr. D. L. Janecek Mr. C. E. Brooks
- Mr. M. L. Jones Mr. F. G. Burford . Mr. R. J. King (RB - LIC)
Mr. R. W. Byrd Mr. C. W. Lambert Mr. W. C. Cade Mr. M. J. Larson Mr. R. W. Carrol Ms. J. M. Manzella (W3 - QP) ;
Ms. T. M. Carter ** Mr. J. R. McGaha (ECH) !
Mr. W. F. Clark Mr. T. O. McIntyre Mr. D. G. Cupstid Mr. R. V. Moomaw Mr. J. P. Czaika Mr. J. E. Owens ,
Mr. L. F. Daughtery Mr. S. D. Reeves
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Mr. G. G. Davie (W3 - LIC) Mr. J. C. Roberts !
Mr. D. R. Denton (ANO) Mr. J. L. Robertson I Mr. M. A. Dietrich (RB - QP) Mr. C. D. Stafford j
- Mr. C. M. Dugger (W3 - VP) Mr. G. D. Swords j Mr. W. A. Eaton Ms. E. A. Turnage Mr. C. W. Elissesser Mr. R. L. Thomas (ECH)
Mr. J. L. Ensley (ESl) Mr. T. H. Thurmon Mr. E. C. Ewing (W3 - LIC) Mr. J. D. Vandergrift (ANO - LIC)
Mr. R. E. Garrison (M-ECH-411) Mr. J. E. Venable Mr. T. J. Gaudet (W3 - LIC) Mr. A. D. Wells (RB)
Ms. C. W. Gunn Mr. R. A. Wilson
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Attachment I to GNRO-99/00049 BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION l l
! LICENSE NO. NPF-29 l
DOCKET NO. 50-416 l
IN THE MATTER OF ENTERGY MISSISSIPPI,INC.
and SYSTEM ENERGY RESOURCES,INC.
and SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION and ENTERGY OPERATIONS,INC.
AFFIRMATION I, W. A. Eaton, being duly sworn, state that I am Vice President, Operations 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; that I signed this application as Vice President, Operations GGNS 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.
Nh W. A.' Eatori STATE OF MISSISSIPPI COUNTY OF CLAIBORNE SUBSCRIBED AND SWORN TO before me, a Notary Public, in and for the County and State above named, this d2" day of %e ,4998/75y w
' (SEAL)
Notarf Public 1
Mt3BISSIPPI STATEWiBElI0TARY PWUC i l
g-4 TMU k $Y L.
Attachment 2 to GNRO-99/00049 l
ATTACHMENT 2 PROPOSED CIIANGE TO TIIE OPERATING LICENSE 1
FUEL IIANDLING ACCIDENT OPERATIONAL CONDITIONS DISCUSSION of PROPOSED CIIANGES l
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Attachment 2 to GNRo-99/00049 i
AFFECTED TECHNICAL SPECIFICATIONS l The following are the Technical Specifications and License Conditions affected by the l proposed change. !
Operating License Condition Affected pages 2.C(2) 4 and 5 Limiting Condition for Operation (LCO) Affected pages 3.3.6.1 Primary Containment and DrywellIsolation 3.3-51,3.3.-52, and 3.3-55 Instrumentation 3.3.6.2 Secondary Containment Isolation 3.3-62 Instrumentation 3.6.1.3 Primary Containment Isolation Valves 3.6-13 (PCIVs) 3.6.4.1 Secondary Containment 3.6-42 and 3.6-43 3.6.4.2 Secondary Containment Isolation Valves 3.6-45 and 3.6-47 (SCIVs) 3.6.4.3 Standby Gas Treatment (SGT) System 3.6-49 and 3.6-50 Associated Technical Specification Bases changes to be implemented following NRC approval of the proposed Technical Specification changes are detailed in Attaclunent 3.
SCOPE This proposed amendment to the GGNS Technical Specifications requests revisions to selected specifications that imposed certain restrictions applicable during Modes 4 or 5.
The affected Limiting Conditions for Operation (LCOs) have been evaluated and it has been determined that the shutdown safety controls associated with handling irradiated fuel in the primary or secondary containment and performing CORE ALTERATIONS ;
may be relaxed. The controls on these systems and components are not required to limit i doses for the full time that these special operations may be taking place. The purpose of the change is to revise the OPERABILITY requirements for mitigating systems during Modes 4 and 5. During these modes, the limiting accident is the Fuel Handling Accident (FHA). One aspect of the change is to better defime the period during which operability of those systems that mitigate the consequences of an FHA is actually required in order to meet the applicable dose acceptance criteria. Currently, this acceptance criteria is 75 rem thyroid,6 rem whole body for offsite doses (i.e.,25% of 10CFR100 limits) and 5 rem whole body for control room doses). A second aspect deletes the constraint on operations during CORE ALTERATIONS.
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t, Attachment 2 to GNRo-99/00049
SUMMARY
OF CHANGES The relaxation of the shutdown safety controls involves changes to the following Operating License condition and Technical Specifications:
' 1. Technical Specification 3.3.6.1 - Primary Containment and Drywell Isolation Instrumentation
- 2. Technical Specification 3.3.6.2 - Secondary Containment Isolation Instrumentation
- 3. Technical Specification 3.6.1.3 - PCIVs (due to proposed Applicability of 3.3.6.1,
' Function 2.g.)
- 6. Technical Specification 3.6.4.3 - SGT System 7.- Operating License Condition 2.C(2)
The proposed changes are summarized below.
- 1) The Applicability Statements for each of the above LCOs (except 3.6.1.3) are proposed to be modified from "when handling irradiated fuel assemblies" to "when handling recently irradiated fuel assemblies". Also, revised wording of both the Conditions and Required Actions are proposed to be consistent with the change in the LCO Applicability Statement. Note that a markup of the wording
- of the Applicability Statement for LCO 3.6.1.3 is not included in this package.
There is no written change required; the applicability of this specification is affected by its reference to LCO 3.3.6.1 Function 2.g, which is being revised per this item. The net result of this proposal is to establish a new term for the irradiated fuel that contains sufficient fission products to require the operability of accident mitiption systems to meet the accident analysis assumptions. This new term is then used to establish operational conditions where specific activities may represent situations where significant radioactive releases can be postulated and to redefine the appropriate operability requirements for the associated ESF systems.
The actual definition of the term recently irradiated fuel assemblies will be included in the Bases for each of these specifications and is described further in the Discussion section below.
The use of the term "recently irradiated fuel" provides a mechanism for applying a cutoffin fission product decay to the various specifications where the concept is applied. The actual definition of this term will be included in the Bases for each of the affected Technical Specifications. Based on the results of a recent revision to the analysis of the Fuel Handling Accident, GGNS has determined that an 8-day decay period is sufficient to assure that the radiological consequences are
- within the acceptance criteria of SRP 15.7.4 [Ref.15] and 10CFR50, Appendix A,
' General Design Criteria 19 (Ref.16].
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Attachment 2 to GNRO-99/00049
- 2) The Applicability Statements for the above LCOs are also proposed to be
. modified to no longer require the LCO be met during CORE ALTERATIONS.
Revised wording of both the Conditions and Required /.ctions are proposed to be consistent with the change in the Applicability Statement. (Note - as in item 1 above, there is no revision required to the LCO 3.6.1.3 Applicability Statement; it is included here to highlight the impact of revising the referenced LCO 3.3.6.1 Function 2.g.)
. In addition to fuel handling accident (FHA), the accidents postulated to occur during core alterations are described in the UFSAR [Ref.17, see Section 15.4].
They include the control rod withdrawai error and the misplaced fuel bundle (i.e.,
inadvertent loading and operation of a fuel assembly in an improper location) everts. These events were determined to not involve any 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 limiting FHA event, involving a recently irradiated fuel assembly dropped onto other recently irradiated fuel assemblies in the storage racks, is the basis for the operability constraints being imposed on the mitigating systems in item 1 above.
Considering those constraints, the only FHA that could occur as a part of core alterations would be the drop of an unirrradiated assembly on the recently irradiated assemblies in the reactor. This event has also been analyzed and the dose consequences of this event were found to meet the acceptance criteria. Thus the proposed Technical Specification requirement changes deleting the CORE ALTERATIONS constraint is acceptable. The LCO Applicability Statements related to operations with a potential for draining the reactor vessel are unaffected by the proposed changes.
- 3) It is also requested that the words appended to Operating License Condition 2.C(2) describing the one time only allowance associated with this same issue be deleted. This one time only allowance was granted by Amendment 129 to the Operating License, was applicable during the GGNS Refueling Outage RF08, and is no longer required. This is an administrative change.
DISCUSSION Following reactor shutdown, decay of the short lived fission products greatly reduces the fission product inventory present in irradiated fuel. Changes are being proposed to the Technical Specifications noted above to redefine the operability requirements for selected engineered safety feature (ESF) systems. The proposed changes are based on an analysis (Reference 3) of the design basis accident postulated to involve a fuel cladding breach during core alterations or Modes 4 and 5. This analysis demonstrates the radionuclide inventory has sufficiently decreased after 8 days to assure that the consequences are within the applicable dose acceptance criteria. This analysis takes no credit for either primary or secondary containment isolation nor for operation of the Standby Gas Treatment System. The Control Room Fresh Air System has been credited in order to 3
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Attachment 2 to GNRO-99/00049 assure control room doses remain within limits. On this basis, the existing shutdown safety controls in the affected Technical Specifications are not required.
Implementation of the proposed changes will have a significant impact on outage activities at Grand Gulf Nuclear Station (GGNS) and are expected to result in reduced outage costs and increased flexibility with no impact on safety margin. Currently, moving large equipment into secondary containment such as chemical-decontamination equipment or safety-relief valves must either be delayed or moved through an alternate entrance. Additionally, the high level of modification, maintenance, and repair activities during outages, in conjunction with the existing constraints, can create access problems.
This situation, in turn, may result in less effective approaches to the work activity, less convenient access paths, less efficient utilization of personnel and contract resources, and limit the application of the best solutions to emergent issues. Also, additional door guards are typically employed during outages to' ensure that incidents of doors being left open are minimized. For the upcoming outage alone, the requested changes are predicted to permit a net one day savings in the overall outage schedule. This day is conservatively valued at over $500,000. The increased schedule flexibility and the better resource utilization permitted by this change are further expected to result in accrued cost reductionk well in excess of another $500,000 over the remaining operating life of the plant.
Current Licensing Basis The current GGNS Licensing Basis includes the consideration of two Fuel Handling Accidents. The Fuel Handling Accident in the auxiliary building is evaluated in the GGNS Updated Final Safety Analysis Report (UFSAR) Section 15.7.4 and that in the ;
containment building is described in Section 15.7.6. The design basis analyses of both of I these events are based on the Standard Review Plan (SRP) 15.7.4 and Regulatory Guide (RG) 1.25. The evaluation of the FHA in the containment considers two cases: 1) activity released from the fuel is conservatively assumed to be completely pulled through {
the open equipment hatch into the auxiliary building; and 2) activity is released partially i into the auxiliary building and partially through the containment ventilation system. The results of both of these events demonstrated that offsite doses were well within both 10CFR100.11 limits and Standard Review Plan 15.7.4 limits (75 rem thyroid and 6 rem whole body).
For the above fuel handling accidents, secondary containment integrity, isolation of the
- containment and fuel handling area ventilation systems, and the Standby Gas Treatment :
System (SGTS) limit the transport of fission products to the environment and the I associated radiological consequences (per SRP 15.7.4 guidelines) to well within the 1
10CFR100.11 limits. 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 I handling accident, appropriate operating restrictions are imposed by the Technical Specifications, The Technical Specifications define operability, 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 1:
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Attachment 2 to GNRo-99/00049 i
damper / valves and the associated electrical power systems. These systems limit the l transport of fission products to the environment such that the radiological effects at the l Site Boundary are approximately 1.31 rem whole body and 8.67 rem thyroid [ Reference l 17, see Tables 15.7-10 and 15.7-14).
' Reanalysis of Fuel Handling Accident EOI has reanalyzed the Fuel Handling Accidents for GGNS [ Reference 3]. The reanalysis was performed to:
e incorporate Federal Guidance Report (FGR) 11 and 12 dose conversion i factors, e update atmospheric dispersion factors (x/Q factors), and j e- assess the impact of not crediting various engineered safety feature (ESP)
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systems currently credited in reducing the consequences of the FHA. !
Precursors for this event are unchanged from those described in the UFSAR; however,
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the analysis was expanded to evaluate the effects of various decay time periods in j conjunction with the assumption that those systems used to mitigate the accident as j described in the UFSAR are not available. 1 The FGR 11 and 12 dose conversion factors are based on additional empirical information and improved understanding of radiation effects. Atmospheric dispersion J values (x/Q factors) are based on meteorological conditions in the area surrounding the j I
site. GGNS has been gathering this data since before the plant began operation and has updated the various factors based on that data. This recent update applied the data from the previous 6 years (1992-1997). The offsite dispersion values were developed with the PAVAN code documented in NUREG/CR-2858. The control room dispersion values were based on the closest release point to the control room intake and developed with the ARCON96 code documented in NUREG/CR-6331, Rev.1.
The analysis demonstrates that for the worst case drop, the resulting dose consequences are within the limits established in SRP 15.7.4 (offsite) and GDC 19 (control room). This conclusion considers both the period (first 8 days after shutdown) when mitigating systems are required and the subsequent period when no credit has been taken for primary or secondary containment integrity nor for the operation of the Standby Gas Treatment System. A confirmatory fuel handling accident analysis was performed by the NRC Staffin support of the approval of Amendment 129 to the GGNS Operating License (Reference 18). Note that the earlier request had required a longer decay period (twelve days); the switch to the revised atmospheric dispersion factors permits the reduction in the required decay period.
Key assumptions used in the analysis are as follows:
o Regulatory Guide 1.25 [Ref.19] assumptions are followed with the exception that the FGR 11 and 12 dose conversion factors are used for thyroid dose and whole body dose calculations.
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e Attachment 2 to GNRo-99/00049 o Consistent with the results provided in NUREG/CR-5009, a release fraction of 12% was applied to 1-131 for extended burnup fuel.
o It is conservatively assumed that all sections (i.e.,3",4", and 5") of the NF500 telescoping mast and the handling tool (assumed weigh + of 619 pounds) are attached to the dropped assembly.
o Credit is taken for the buoyancy force on the dropped object. No credit is taken for drag force on the dropped object.
o All fuel rods of a dropped bundle are assumed to fail due to bending (i.e., no credit is taken for lateral support provided by a fuel channel).
o . Per RG 1.25, 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 as follows:
10% of the noble gases (excluding Kr-85) 30% of the Kr-85 inventory 10% of the Iodine inventory (excluding I-131) 12% of the 1-131 inventory o Per RG 1.25, the fission products released to the containment / auxiliary building escape to the environment within two hours.
o No credit is taken for the iodine removal capability of the containment
. ventilation system charcoal filters.
o Key inputs to the analysis include:
- No credit for containment integrity (direct release to the environment)
Normal control room ventilation and filtration No credit for Standby Gas Treatment System Power level 3910 MWth Radial Peaking Factor 1.7 Radioactive Decay Period 8 days Fuel rods damaged 98 Total # of rods in core 56,000 Gap iodine chemical form 0.0025 organic 0.9975 inorganic 3
Control room x/Q (0-2 hours) 2.75E-03 s/m 3
Offsite x/Q (0-2 hours) 9.56E-04 s/m 3
Control room volume 2.53E+05 ft Control room infiltration 600 cfm Control room recirculation flow through charcoal bed 4,000 cfm Control room charcoal iodine removal efficiency 95 %
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( Attachment 2 to GNRO-99/00019 i
o Offsite 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. Control room atmospheric dispersion factors were developed using the NRC ARCON96 code.
o Per RG 1.25, the activity released to the containment / auxiliary building is I based on an overall decontamination factor of 100 for Iodine for the first 23 feet of water coverage (additional decontamination for water depths of 46 feet l or greater is credited) and a decentamination factor of one for the noble gases L (i.e., no noble gases are retained in the pool).
Supplemental Shutdown Risk Justification
- i. The containment and associated engineered safety feature systems are only required by l the Technical Specifications during the specific activities which are postulated to result in I
- a significant release of radioactivity (e.g., fuel handling accident, drain down). As a
! result, the requirements of the Technical Specifications are based on the plant being in l
specified conditions and are not based on providing requirements associated with shutdown risk considerations. Shutdown risk issues are addressed by utility outage management which follows the guidance of NUMARC 91-06, Guidelines for Industry ;
Actions To Assess Shutdown Management [Ref. 20]. NUMARC 91-06 Section 4.5 discusses the need to assure that secondary containment, for GGNS, closure can be achieved to prevent fission product release during severe accidents. NUMARC 91-06
- also identifies that the time to effect closure should be consistent with plant conditions l (e.g., reactor coolant system inventory and decay heat load). Consistent with the i
industry's commitment in the letter from NUMARC's President, Mr. Byron Lee, Jr., to Mr. James M. Taylor of the NRC [Ref. 21], GGNS has administrative controls in place to meet the recommendations of NUMARC 91-06 Section 4.5 for extended loss of decay heat removal events.
In the draft NUMARC~93 01 guideline, Section 11.2.6.5," Safety Assessment for n Removal of Equipment from Service During Shutdown Conditions," under the
- subheading of" Containment - Primary (PWR)/ Secondary (BWR)", the following guidance is provided.
"... forplants which obtain amendments to modify Technical Specification requirements on primary or secondary containment operability and ventilation system operability duringfuel handling or core alterations, thefollowing guidelines should be included in the assessment ofsystems removedfrom service:
e Duringfuel handling / core alterations, ventilation system and radiation monitor availability (as defined in NUniRC 91-06) should be assessed, with respect to ,
filtration and monitoring ofreleasesfrom thefuel. Following shutdown, i radioactivity in thefuel decays awayfairly rapidly. The basis ofthe Technical ,
Speciflcation operability amendment is the reduction in doses due to such decay. i The goal ofmaintaining ventilation system and radiation monitor availability is to '
reduce doses evenfurther below thatprovided by the natural decay.
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Attachment 2 to GNRo-99/00049
- A single normal or contingency method to promptly close primary or secondary containmentpenetrations should be developed. Such prompt methods need not completely block the penetration or be capable ofresistingpressure. "
The nurpose of the " prompt methods" mentioned above is to enable ventilation systems )
to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored.
In the interim period until the revision to NUMARC 93-01 is endorsed as a formal industry position, GGNS will adopt these provisions for controlling the removal from service of systems, structures and components (SSC's) that are currently required by Technical Specifications during core alteration / fuel handling periods.
Also, in accordance with Technical Specification 3.9.6, RPV Water Level - Irradiated Fuel, handling irradiated fuel in the reactor vessel can only occur when the water level in the reactor cavity is at the high water level. Thus, the proposed changes only affect containment requirements during relatively low risk times during refueling outages.
Therefore, the proposed changes do not significantly increase the shutdown risk.
Additionally, the proposed Technical Specification changes do not affect the requirements to have the containment systems available any time the unit is in MODE 1, 2, or 3 regardless of whether fuel handling is occurring in the spent fuel pool.
The impact of the proposed changes on the public health risk profiles was also evaluated using the GGNS probabilistic safety assessment (PSA) models. The conclusions from this evaluation are that the public health risk impact in terms of the Core Damage Frequency (CDF) and the Large Early Release Frequency (LERF) is negligible. ,
This change does not impact the GGNS ORAM calculations of risk metrics (coce damage risk and boiling risk). ORAM does not calculate the Large Early Release Frequency (LERF) risk profile. Of those accidents during Modes 4 and 5 which are postulated to c
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result in a release, the fuel handling accident produces a small release and the loss of shutdown cooling event is a much more slowly evolving scenario that allows evacuation prior to release. Therefore, the LERF profile during this operation is essentially zero.
CONSISTENCY WITH RELATED INDUSTRY SUBMITTALS EOI has been working with the Staff, on both a plant-specific and an industry-wide i generic basis, since November 1994 to resolve issues associated with the proposed change [ References 4-13]. One of the groups EOI has been working with is the IlWR6 group (i.e., Clinton, Grand Gulf, Perry, River Bend). A generic technical specification change designated Technical Specification Traveler Form (TSTF)-51 [ Reference 14] has been submitted for NRC review. As of a 12/15/98 status report, the NRC was continuing to discuss TSTF-51. The attached proposed changes are consistent with those proposed generically under the TSTF. The major difference between the GGNS and the generic 8
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l Attachment 2 to GNRo-99/00049 l proposed changes is the number of affected Technical Specifications. The GGNS l analysis of the Fuel Handling Accident credits the operation of the Control Room Fresh Air System; as a result, relaxation of the fuel movement operability requirement for the CRFA and the CR AC systems, as well as the supporting electrical power systems has been retained.-
1 In addition, the Perry Nuclear Power Plant (PNPP) plant-specific submittal has now been approved by Amendment 102 to their operating license [Refen nce 22]. The attached l proposed changes are consistent with those approved for PNPP for the equivalent specifications.- PNPP also changed several specifications that do not exist at GGNS.
And, as discussed above, the Control Room Fresh Air and Air Conditioning Systems and electrical power system operability requirements during fuel movements and core alterations have been retained for GGNS.
Additionally, the requested changes are similar in nature to those approved on a one-time basis in Amendment 129 to the GGNS Operating License [ Reference 10].
REFERENCES
- 1) U.S. Atomic Energy Commission;" Calculation of Distance Factors For Power And Test Reactor Sites", TID-14844, March 1962, 1
- 2) GGNS Calculation XC-Q1111-98011, Revision 0," Control Room x/Q Analysis."
- 3) GGNS Calculation XC-QlJll-96005, Revision 1," Design Basis Fuel Handling Accident."
- 4) Letter GNRO-94/00131, Fuel Handling Accident Operational Conditions, Proposed f Amendment to the Operating License (PCOL-93/08), dated November 9,1994.
- 5) Letter GNRO-95/00090, Fuel Handling Accident Operational Conditions, Proposed Amendment to the Operating License (PCOL-93/08 Revision 1), dated August 4, 1995.
- 6) Letter GNRI-95/00158, Summary of Meeting on July 20,1995, regarding Systems required while handlirg irradiated fuel in containment, dated August 25,1995.
- 7) Letter GNRO 96/00068, Enclosure Building Isolation Requirements during Refueling l Outage 8, Proposed Amendment to the Operating License, dated June 20,1996.
- 8) Letter GNRO-96/00105, Enclosure Building Isolation Requirements during Refueling )
Outage 8, Additional Information September 11,1996. ;
- 9) Letter GNRO-96/00048, Fuel Handling Accident Operational Conditions, Proposed Amendment to the Operating License Additional Information, dated April 24,1996.
9 l
m, Attachment 2 to GNRo-9%0049
[w l 10) Letter GNRI-96/00213, Issuance of Amendment No.129 to Facility Operating
- License No. NPF Grand Gulf Nuclear Station, Unit 1 (TAC No. M95762), dated October 18,1996.
I 1) Letter GNRI-96/00119, Fuel Handling Operational Conditions (TAC No. M91229),
dated May 24,1996.
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- 12) Letter GNRI-96/00128 Meeting on Defense-in-Depth to mitigate the fuel handling Accident, dated June 4,1996.
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- 13) Letter GNRI-96/00148, Meeting on Enclosure Building Isolation Requirements during refueling outage number 8 (TAC No. M95762), dated July 9,1996.
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- 14) Technical Specification Traveler Form-51, Revision 1, " Revise containment requirements during handling irradiated fuel and core alterations," April 14,1999.
- 15)NUREG-0800, Standard Review Plan, Section 15.7.4, " Radiological Consequences of Fuel Handling Accidents," Revision 1, July 1981.
16)10CFR50 17)GGNS UFSAR, various sections ,
- 18) GGNS Operating License, as amended through Amendment 137, dated May 17, 1999. ,
i i
- 19) Regulatory Guide 1.25," Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," dated March,1972. i 20)NUMARC 91-06, Guidelines For Industry Actions To Assess Shutdown Management, December 1991,
- 21) Letter from Mr. Byron Lee, Jr., President and Chief Executise Officer NUMARC, to Mr. James M. Taylor, Executive Director for Operations U.S. NRC, dated December ,
6,1991. !
- 22) Letter, NRC to L. W. Myers, " Amendment No.102 to Facility Operating License No.
NPF Perry Nuclear Power Plant, Unit 1 (TAC No. M94028)", dated March 11, 1999.
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.k Attachment 2 to GNRo-99/00049 P
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 uses a new term to describe irradiated fuel that contains sufficient fission products to require operability of accident mitigation sysk ns to meet the accident analysis assumptions and deletes the CORE ALTERATIONS constraint. This proposed change revises the operability requirements and required actions for the following Technical Specification Limiting Conditions for Operation (LCOs): Primary Containment and Drywell Isolation Instrumentation, Secondary Containment Isolation Instrumentation, PCIVs, Secondary Containment, SCIVs, and SGT System. In addition to the requested changes to the Technical Specifications, the License Condition describing the one time only allowance associated with this same issue is requested to be
. deleted. This one time only allowance was granted by Amendment 129 to the Operating License and is no longer in affect. For accuracy and completeness, this obsolete
' allowance is requested to be deleted as part of this change.
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 ifoperation 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. I A new term to describe irradiated fuel is used to establish operational conditions where i specific activitics represent situations where significant radioactive releases can be 1 postulated. These operational conditions are consistent with the design basis analysis.
Because the equipment affected by the revised operational conditions is not considered an l initiator to any previously analyzed accident, inoperability of the equipment cannot increase the probability of any previously evaluated accident. The proposed requirements bound 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 ;
l evaluated accident. ;
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Attachment 2 to GNRo-99/00049 Removing a one time only allowance granted by Amendment 129 to the Operating License that is no longer in affect is an administrative change. Therefore, the proposed change does not significantly increase the probability or consequences of any previously .
evaluated accident. l Based on the above, neither the proposed changes to the Technical Specifications nor that to the Operating License 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 new term to describe irradiated fuel 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.
Removing a one time only allowance granted by Amendment 129 to the Operating License that is no longer in affect is an administrative change. Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previous analyzed.
- Based on the above, neither the proposed changes to the Technical Specifications nor that to the Operating License create the possibility of a new or different kind of accident from any accident previously analyzed.
- 3. The proposed changes do not involve a significant reduction in a margin of safety.
The new term to describe irradiated fuel 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 and are established such 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 proposed change only eliminates the l unnecessary 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 10CFR100 limits and the licensing limit defined by NUREG 0800, Section 15.7.4. I With respect to the control room personnel doses, the margin of safety is the difference l between the 10CFR100 limits and the licensing limit defined by 10CFR50, Appendix A, l Criterion 19 (GDC 19). The additional margin between the calculated doses for the !
postulated events and the corresponding licensing limit provides no useful purpose. i l
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a N.
., Attachment 2 to GNRO-99/00049 1
i The proposed applicability continues to ensure that the whole-body and thyroid doses at l both the control room and the exclusion area and low population zone boundaries 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.
Removing a one time only allowance granted by Amendment 129 to the Operating !
License that is no longer in affect is an administrative change. Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, neither the proposed changes to the Technical Specifications nor that to the Operating License 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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Attachment 3 to GNRO-99/00049 1
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ATTACHMENT 3 i
I PROPOSED CHANGE TO TIIE OPERATING LICENSE f 1
FUEL IIANDLING ACCIDENT OPERATIONAL CONDITIONS MARKED UP PAGES I l
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4 (b) SERI is required to notify the NRC in writing prior to any change in (i) the terms or conditions of any new or existing sale or lease agreements executed as part of the above
~
authorized financial transactions, -(11) the GGNS Unit 1 operating agreement, (iii) the existing property insurance coverage for GGNS l Unit 1 that would materially alter the representations and conditions set forth in the Staff's Safety Evaluation Report dated December 19, 1988 attached to Amendment j No. 54. In addition, SERI id required to j i . notify the NRC of any action by a lessor or l
other successor-in interest to SERI that may -
have an effect on the operation of the facility.
C. The license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions !
specified or incorporated below:
L (1) wav4=u= Dn==v Taval Entergy Operations, Inc. is authorised to operate the facility at reactor core power levels not in exC3ss of 3833 megawatts thermal (100 percent power) in accordance with the conditions specified herein.
(2) Tarhn4r=1 anar474r=*4mn=
The Technical Specifications contained in Appendix A
.and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 136 are hereby incorporated into this license. Entergy Operations, Inc.~ shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. J
=at-that the enclosure building may be inoperable
, during TIONS and movement of non-recently
' irradiated fuel (i.e., at has not. occupied part l l uring RFO 8 of a critical reactor core for l
. . and .tbet._ standby gas treatment (SGT) system l l !
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4 Amendment 136 l
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a unable to automat y start or achieve and maintain the required vacuum, rovided the following conditions exist:
. All dampers c m sicating tween the auxiliary 11 ding and the enclosure buih i are closed.
- b. The ac e door between the auxiliary bu ng and the e osure building is closed, except when the acce opening is being used for entry and exit.
The SGT system is blocke rom automatic itiation.
- d. SGT syste available for manual in'- ation or the ACTIONS of .6.4.3 are complied w .
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I da Amendment 129