ML20195B243

Requests NRC Review & Approval of Rev to GGNS Licensing Basis,Using One of Revised Source Term Insights Developed & Described in NUREG-1465.Detailed Discussion of Changes,Encl

ML20195B243
Person / Time
Site:	Grand Gulf
Issue date:	11/03/1998
From:	Hughey W ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-1465 GNRO-98-00085, GNRO-98-85, NUDOCS 9811160072
Download: ML20195B243 (20)

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O Ent y rstions, inc.

Port Gibson. MS 39150 Tel 601437-6470 W.K.Hughey ar Safety & Reguttory

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- November 3,1998 U.S. Nuclear Regulatory Commission Mail Station P1-37 Washington, D.C. '.0555 Attention:

Document Control Desk

Subject:

GGNS Pilot Application Submittal of the NUREG-1465 Revised Source Term insights Docket No. 50-416 License No. NPF-29 GNRO-98/00085 Gentlemen:

This letter provides for NRC Staff review and approval a request to revise the GGNS licensing basis. The requested change makes use of one of the revised source term insights developed and described in NUREG-1465. This represents a limited scope application of the NUREG findings in that the change takes credit for only one of the insights; in this case, the insight related to the timing and duration of the fission product release. Grand Gulf is a pilot plant in the collaborative efforts of the Nuclear Regulatory Commission (NRC), Nuclear Energy Institute (NEI), and the Electric Power Research Institute (EPRI) for the implementation of the NRC research efforts documented in the NUREG.

- A detailed discussion of the requested change is included in the attachment. In summary, Grand Gulf proposes to revise the LOCA accident analysis scenario to recognize that the initial radioactive release will consist of reactor coolant only.

There will be no instantaneous release of fission products due to fuel failure as is assumed in the current analysis methodology performed in accordance with TID-14844. Because this change makes use of insights that are currently not approved

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by the NRC Staff, GGNS requests the NRC issue a Safety Evaluation Report approving the utilization of the NUREG-1465 findings regarding the initial coolant release phase of an accident.

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GNRO 98/00085 l

Page 2 of 2 The requested change is also based on a generic BWR analysis performed by l

General Electric. The analysis was performed to better model BWR phenomenology in support of industry efforts related to NUREG 1465, which derived much of its findings from PWR-based studies. A report describing the analysis effort, " Prediction of the Onset of Fission Gas Release from Fuel in a l

Generic BWR", was previously submitted to the NRC under the GGNS docket by letter dated May 6,1997 (GNRO 97-0034). This report presented the conservative estimate of the duration of the coolant release phase of the accident used in this submittal.

GGNS is also currently preparing analytical support for a future submittal, which will request approval for the complete elimination of the TID-14844 methodology from our licensing basis dose analysis in favor of that described in NUREG-1465.

GGNS expects to submit this full scope application of revised source term during the first quarter of 1999.

I We appreciate the opportunity to participate in the NUREG-1465 evaluation efforts both as a pilot application plant and as a rebaselining analysis subject. We look forward to continued cooperation on this project as the NRC Rulemaking Plan is implemented. We support the revised source term initiative and believe it is an i

important step toward risk-informed regulatory policy. If you have any questions regarding this submittal, please contact Jerry Burford at 601-437-2714.

l Yours truly,

/FGB attachment:

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)

Regional Administrator i

U.S. Nuclear Regulatory Commission Region IV l

611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 I

Mr. J. N. Donohew, Project Manager (w/2)

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Proposed Change in Primary Containment Isolation Valve Maximum Isolation Time based on the Revised Source Term Purpose Grand Gulf Nuclear Station (GGNS) proposes to revise the licensing basis for the release of fission products following an accident. One of the insights established in NUREG-1465 is that there is a delay in the release of radioactivity from the fuel. Approval of this change will allow GGNS to increase the containment isolation valve closure times credited for limiting post-accident doses to both control room personnel and to offsite individuals. While this new basis would be applicable to all of the containment isolation valves, it only addresses the dose mitigation aspects of the closure requirements. There are currently some valves for which the closure time is limited based on other functional performance requirements (e.g., line break isolation). This submittal does not propose any changes that would eliminate any of these other requirements. The allowable closure times for these valves would not be affected by this proposed change.

This proposed change essentially establishes and documents a quantifiable design basis for the isolation valve closure time requirement. Extremely conservative assumptions for the post accident source term derived from TID-14844 had resulted in regulatory I

guidance for isolation valve closure times to be "less than 60 seconds." Hoivever, new insights regarding the source term for the design basis accident (DBA) scenario recognize that radioactive fission product releases into containment during a DBA Loss of Coolant Accident (LOCA) are not instantaneous. This conclusion was developed in NUREG-1465, Accident Source Terms fhr Light-Water Nuclear Power Plants (ref.1). While NUREG-1465 identified four areas in which development of the source term data may be more realistic than that described in TID-14844, this submittal is requesting a limited-scope application of the revised source term insights. Only the insight related to the timing and duration of the radioactive release is credited here. The duration of the coolant blowdown phase and the timing of the fission product release constitute the licensing basis change regarding the isolation of containment requested here. This change will support an increase in the associated primary containment isolation valve (PCIV) closure time limits.

Grand Gulf Nuclear Station (GGNS) is a pilot plant in the effort to evaluate the potential j

application of the insights regarding the revised source term. In fact, GGNS is also the BWR evaluated by the NRC in the revised source term rebaselining effort using the i

NUREG-1465 insights. The results of this recently completed evaluation are documented in Reference 11. As noted above, this submittal represents a limited scope application of a single aspect of the new source term methodology. GGNS is currently preparing analytical support for a future submittal, which will request approval for the complete elimination of the TID-14844 methodology from our licensing basis dose analysis in i

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favor of that described in NUREG-1465. GGNS expects to submit this full scope application of revised source term during the first quarter of 1999.

l This submittal is presented for NRC review and approval consistent with the intent of the objectives of the pilot program and agreements made between the NRC and the pilot program licensees. As a part of the pilot program, the requested change has not been subject to a 10CFR50.59 review; it is being conservatively submitted similar to an unreviewed safety question. While the revised source term results in lower dose consequences, particularly for the time period ofinterest in this request, the NRC is still evaluating the methodology and results. Further, this request is based in part on a BWROG report that has not yet been approved by the NRC. It is requested the NRC issue a Safety Evaluation Report approving the utilization of the NUREG-1465 insights.

Background

The Atomic Energy Commission published TID-14844," Calculation of Distance Factors l

for Power and Test Reactors," in 1962. This document established conservative accident source terms to be used in the evaluation of the dose consequences of design basis accidents. Indeed, use of this information in developing the plant safety analyses was encouraged in both Regulatory Guides and the Standard Review Plan to deruonstrate compliance with 10CFR100.

Research and operating experience over the past 30 years has expanded the understanding of severe accidents and the behavior of fission product releases. NUREG-1465,

" Accident Source Terms for Light-Water Nuclear Power Plants," was published in February 1995 with a primary objective of defining revised accident source terms for use in the licensing of future Light Water Reactors (LWRs). The DBA source terms developed in the NUREG are similar to those in TID-14844, but recognize a more realistic scenario of release timing, composition, chemical form, and removal mechanisms.

Since the NUREG source term data is expected to result in lower calculated dose consequences, the NRC has decided not to require current plants to revise their accident analyses. The Advisory Committee on Reactor Safeguards (ACRS) has encouraged the use of the revised source term. Considering the role of accident analyses and dose consequences in establishing the design and operating parameters for safety related systems, the revised source terms create the potential for new cost beneficial changes which maintain an acceptable level of safety. The NRC is currently developing rulemaking, which would establish a regulatory framework for existing plants to make use of the revised source terms. The Nuclear Energy Institute (NEI) has coordinated an industry initiative to propose a suitable methodology for the use of the revised source terms. In conjunction with the Electric Power Research Institute (EPRI), NEI developed l

Technical Report (TR) 105909, " Generic Framework for Application of Revised 4

Accident Source Term to Operating Plants"(ref. 3). It was submitted to the NRC in November 1995.

i The EPRI document outlines two methodologies for the application of the revised source terms. One outlines a process involving an extensive revision of the plant's safety analyses to model all or most of the new source term insights in the NUREG (i.e., timing, 1

chemical form, and release and removal mechanisms). The second method would permit the use of selected insights from the NUREG in limited scope applications. This second approach could be used by licensees without the extensive design and licensing efTorts involved in the development of a full scope application. The NRC Commissioners recently approved this selective approach in a Staff Requirements Memorandum dated September 4,1998 (ref.12).

l Based on assumptions described in the NUREG, it was recognized that additional analysis was needed to better establish the BWR-specific gap release characteristics. The Boiling Water Reactor Owners Group (BWROG) prepared a conservative analysis determining the minimum time to fuel perforation for a generic BWR following a DB A LOCA with no emergency core cooling system (ECCS) injection. The report, titled

" Prediction of the Onset of Fission Gas Release from Fuel in Generic BWR"(ref. 6) was submitted to the NRC under the GGNS docket number (ref. 7). The values requested in the change to the licensing basis are derived in part from the work in this report. A brief summary of the highlights from these supporting documents is provided below.

Summary of Revised Source Term Insights NUREG-1465 NUREG-1465 represents the culmination of decades of research on fission product release and transport in LWRs under accident conditions. The revised source terms and the transport and removal mechanisms described in the NUREG represent a significant advancement in understanding of accident source terms beyond that described in TID-14844.

The major insights summarized in NUREG-1465 involve the timing and duration of fission product releases, fission product composition and magnitude, chemical form of fission products, and in-containment removal mechanisms.

Timing andDuration The five categories of release phases representing the progress of a severe accident in a light water reactor are described as:

1. Coolant Activity Release

2. Gap Activity Release

3. Early In-Vessel Release

4. Ex-Vessel Release

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S. Late In-Vessel Release i

Phases 1,2, and 3 are considered in current design basis accident (DBA) evaluations, albeit they are all conservatively assumed to occur instantaneously. Phases 4 and 5 are related to severe accident evaluations.

The timing and duration of the release of radioactive material from the core to the containment has been determined using sophisticated computer codes that predict fission product behavior. A more accurate prediction of l

the onset and duration of the coolant activity release is obtained by replacing the instantaneous release of noble gases and iodine isotopes into containment assumed in TID-14844.

l The coolant activity release phase, and thus the time delay prior to the l

_ onset of the gap release of fission products, is an essential basis for this submittal.

Composition andMagnitude The NUREG provides a more accurate definition of the types and quantities of radionuclides released into the containment during postulated accidents to replace the TID-14844 assumptions of 100% of the core inventory of noble gases,50% of the iodine isotopes, and 1% of the remaining solid fission products.

l ChemicalForm The chemical forms ofiodine present in containment are more accurately modeled and predicted in the NUREG-1465 report. The radioactive iodine chemical fonn can greatly affect the efficacy of removal mechanisms and therefore cause significant variation in iodine inventories available for release to the environment.

RemovalMechanisms NUREG-1465 documents an improved understanding of fission product removal mechanisms, which have a fundamental effect on the inventory of fission products in the containment atmosphere available for release to the environment. The effect of engineered safety features (ESFs) and natural r

removal mechanisms are examined.

BWROG Report 4

l The BWROG report (ref. 6) documents the results of an analysis performed to determine the minimum time to the onset of release of radioactive material from perforated fuel assembly following a DBA LOCA. NUREG-1465 assumes a coolant activity phase of 30 seconds based solely on PWR analyses, I

but recognizes that plant specific analyses could justify longer times (see page 8 of ref.1). The BWROG report was commissioned so that the BWR fleet

i would not be unduly penalized by the overly conservative assumptions made in NUREG-1465.

4 The analysis was performed using a limiting plant configuration and fuel type.

NRC-approved codes were used to calculate the minimum duration of the

. coolant activity phase described in the NUREG. The BWR coolant activity release phase, which represents the period from the time of the start of the accident until the initiation of fuel perforation and the attendant gap release, is

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calculated to last 121 seconds.

Details of the Proposed Change Based on the use of NUREG-1465 insights, a more realistic time frame for containment isolation following an accident is appropriate. One such insight from the NUREG recognizes that the entire radioactive source term from a DBA is not immediately released into the drywell, but is characterized by distinct release phases as the accident progresses. Based on these insights, and on the BWR-specific value of the timing of the

. gap activity release phase of a LOCA as calculated in the BWROG report, GGNS proposes to revise its licensing basis. The proposed change replaces the assumption of an instantaneous release of fission products into the drywell with a more accurate scenario in which the gap release is delayed by up to 121-seconds.

.GGNS has performed a site-specific analysis of a release of radioactivity from the reactor coolant only during the first two minutes of an accident and has conservatively assumed that containrnent isolation does not occur during this time. The analysis considers the break flow released to the drywell as flowing into containment through the suppression pool and bypass leakage flowpaths. Because the containment will not be isolated for 121 seconds, all flow into the containment is modeled as released to the environment. The 4

dose associated with the blowdown of reactor coolant from a postulated recirculation line break, assuming the maximum iodine activity in the coolant permitted in the Technical Specifications, has been evaluated for the proposed 121-second containment isolation time and has been determined to be insignificant. This conclusion was also applied in the NRC rebaselining study in which the coolant activity was neglected from the limiting DBA analyses since "it is a small contribution to doses."(Ref. I1) Recognition of the delayed fission product release does not affect the dose consequences of the event as currently presented in UFSAR 15.6.5.5 because the same amount of radioactivity is assumed to be released. That is, the source term defined by TID-14844 is simply released about two minutes later. Other than the two-minute delay, the other existing analysis assumptions remain unchanged.

The conservatism inherent in the old analysis methodology is also supported by the discussion of the LOCA dose analysis provided in the NRC GGNS Safety Evaluation Report, NUREG-0831. Section 15.3.2 includes a description of an independent analysis performed by the staff of the dose consequences of this scenario. The specific values are e

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very conservative and were used to demonstrate compliance with the acceptance limits.

Ofinterest here, is the projected magnitude of the dose reduction due to delaying the fissi.on product release. It is noted:

L Most of the 0- to 2-hour doses are due to the assumption ofinstantmeous

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mixing and transport of the released radioisotopes. If, for example, more than 2 minutes were to elapse before fission products were able to travel i

from the core to the environment through the primary and secondary containment, then 146 rem of thyroid dose and 1.9 rem of whole body dose would be subtracted from the entries in Table 15-1.

[ Table 15-1 presented the results of the independent analysis; the stated l

results are 153 rem thyroid and 3.7 rem whole body. It should be noted l

that the current GGNS UFSAR presents the results of the Entergy LOCA dose consequences analysis in Table 15.6-14.]

A review of the GGNS probabilistic risk assessment (PRA) models indicates that i

l the proposed change to the timing and duration of fission product release assumption made in the GGNS accident analyses does not have any impact on the public health risk profile of GGNS.

l This change involves a limited scope application of the revised source term insights. The timing and duration of the radioactivity release insight requested here involves a change i

to accident dose consequence analysis assumptions that had been made consistent with TID-14844. No other assumptions or methodology changes are proposed. While that l

standard is referenced by 10CFR100,it is referenced as a guide. The dose consequences of this change have been found to be insignificant and LOCA doses are still within 10CFR100 and 10CFR50, App. A, GDC 19 guidelines. No exemption from these j

regulations is required to support this change.

NRC approval of the change to our licensing basis to incorporate the revised source term insight regarding a coolant activity release phase of 121 seconds is requested. The primary change required to the GGNS UFSAR to reflect the change to the licensing basis is a revision to Section 15.6.5.5.2. A markup of that section showing a preliminary version of this change is included in Attachment 4. Once this approval has been granted, GGNS intends to implement the new basis by revising the design basis maximum closure l

time requirements for selected automatic containment isolation valves to 110 seconds.

l This value allows for a 10-second period for starting the diesel generator following an j

accident with a concurrent loss of offsite power. There is also a one-second allowance for l

drywell pressurization and instrument response. This revised closure time requirement assures that the valves will be closed and containmem isolation established within 121 seconds after the accident and prior to any expected release of fission products from the l

fuel assemblies. These changes will be reflected in revisions to the containment isolation descriptions provided in Chapters 6 and 16 of the UFSAR when it is updated as part of the implementation process.

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The actual changes to the design basis closure times for the PCIVs would be handled in l

accordance with the GGNS 10CFR50.59 program requirements. A discussion of the l

potential changes is provided below for your information. The allowable automatic PCIV l

closure times are currently presented in Technical Requirements Manual Table l

TR3.6.1.3-1. These valves and their current closure times are presented in Attachments 1 l

and 2. Attachment 1 includes those valves that currently have allowable closure times of l

60 seconds or greater. Attachment 2 is a list of the valves with analysis-based allowable closure times ofless than 60 seconds or less. The valves affected by this proposed change and that will have their allowable closure times increased to 110 seconds are identified in the attachments.

f The valves affected by the proposed licensing basis change are those automatic primary l

containment isolation valves that have stroke time limits based solely on compliance with 10CFR100 or GDC19 dose limits and for which other specific closure time requirements have not been assumed or considered in the design basis. Note the following:

Selected PCIVs have analysis-based closure times to address system performance a.

requirements, equipment qualification, or other regulatory requirements and will not be affected by this change. Analysis-based closure times are those explicitly assumed or determined in accident analyses and include those valves listed in UFSAR Table 6.2-44, note 'd'.

b. Four PCIVs currently have closure times longer than 110 seconds. E21-F012A and E12-F021B have closure times of 144 seconds based on the physical capabilities of the valve and actuator. Both of these valves are normally closed and isolate penetrations that terminate below the suppression pool surface, communicate with closed systems outside containment, and have significant l

piping runs between the containment and the isolation valves. No additional containment leakage through these penetrations is postulated to occur during the extra 34-second period during which they will remain partially open. E12-F024A-A and E12-F024B-B also currently have an allowable closure time of 144 seconds. These valves are listed in UFSAR Table 6.2-44 note d and thus their closure time has beenjustified by analysis,

c. Drywell and secondary containment isolation valve closure times are not being considered for change at this time.

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d. The containment vent and purge supply and exhaust valves cunently have allowable closure times of 4 seconds. This was based on regulatory guidance that called for isolation valves in direct leak paths to the environment to be capable of L

closing within 5 seconds. However, using the insights of NUREG-1465, the dose consequences oflonger stroke times for these valves are negligible. This was demonstrated by the GGNS analysis of the LOCA event considering all releases from the containment escaping directly into the environment. It is also supported by the existing UFSAR discussion of the fuel handling accident in containment as I

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l presented in Section 15.7.6, which conservatively considers both an open containment hatch as well as purge valve closure times of 120-seconds. As discussed in UFSAR Section 9.4.7.2.2, this delay will not affect the closure of l

these valves due to (i) their locations well away from any debris which might be generated by vent clearing and subsequent suppression pool rise, and (ii) the use i

of debris screens placed on the inner side of each inboard isolation valve.

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e. The eight valves at the end of Attachment 2 are isolation valves for lines from containment sumps to the Auxiliary Building. These lines do not present additional direct leak paths from the drywell. Any drywell connections to these l

lines are provided with redundant drywell isolation valves with allowable closure times ofless than 10-seconds.

An evaluation of the significant hazards considerations and of the environmental impact is provided in Attachment 3.

Potential Benefits of Proposed Change The increase in allowable maximum closure time for PCIVs will result in increased operational flexibility and important safety benefits, some of which are described below:

Valve performance and reliability improvements:

Motor Operated Valves (MOVs) with increased torque switch settings as a result of NRC Generic Letter 89-10 are operating with insufficient margins for weak link l

components and actuator capabilities. These changes have increased the incidence of demand-type failures for marginal valves. With a longer allowable stroke time, the i

actuator can be modified to decrease the overall component stresses and improve both l

margin and reliability. These types of changes are greatly preferred over alternatives such as complete actuator replacement or major valve modifications. Thrust margins can also be improved without modifying the valve.

l Operations improvements:

e The PCIV stroke times are currently tested in accordance with the Technical l

Specifications and the In-Service Testing Program. The acceptance criteria for this l

test are generally the lesser of the maximum allowable closure time per the TRM or the reference stroke time determined per ASME Code guidance.

For some valves, the reference stroke time may be very close to the TRM limit. In such cases, the allowable range over the reference stroke time may be restricted, which increases the potential for failure. (Note - this proposed change to the licensing basis will not influence the reference stroke time of the valves. That value is determined in accordance with Code guidance and is affected by modifying or j

replacing the valve or its actuator.)

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If a valve fails to meet the IST stroke time requirement during surveillance testing, the valve is declared inoperable. This failure results in entry into an Action Statement, the need for a system operability evaluation, and possible preparation of an LER. Increasing the allowable closure times for these valves could decrease the frequency of these events.

System Performance improvements:

The longer stroke times may reduce the potential for water hammer and other detrimental dynamic conditions. This benefit is typically realized for fast-acting valves in steam or water systems. It is not expected to be applicable to the valves j

being considered for increased closure times in this submittal.

Licensing Basis Consistency:

e This change would eliminate the inconsistency between the assumption of an instantaneous release of the fission products and the allowable isolation valve closure times of from 4 to 144 seconds in the FSAR. This assumption was inherent in the non-mechanistic accident source term of TID-14844 and the accepted accident analysis methodology.

References

1) U.S. Nuclear Regulatory Commission; " Accident Source Terms For Light-Water j-Nuclear Power Plants", NUREG-1465, February 1995.

2) U.S. Atomic Energy Commission;" Calculation of Distance Factors For Power And Test Reactor Sites", TID-14844, March 1962.

3) Electric Power Research Institute;" Generic Framework For Application of Revised Source Term To Operating Plants", TR-105909, November 1995.

4) NRC Letter, James M. Taylor to The Commissioners; "Use of the NUREG-1465 Source Term At Operating Plants", SECY-96-242. November 1996

5) NRC Letter, Hugh L. Thompson Jr. to John C. Hoyle, " Staff Requirements-SECY-96-242-Use Of The NUREG-1465 Source Term At Operating Reactors", February l

1997.

6) General Electric Company Report; " Prediction of the Onset of Fission Gas Release from Fuel in Generic BWR", July 1996.

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7) Letter from W.K. Hughey to NRC Document Control Desk; " Submittal of BWROG Report-Prediction of the Onset of Fission Oas Release from Fuel in Generic BWR.

l Application of NUREG-1465 Source Terms for Grand Gulf Nuclear Station L

Rebaselining Study", GNRO-97/034, May 6,1997.

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8) - NRC Letter, L. Joseph Callan to Chairman Jackson et. al.; "Use of Revised Source Term at Operating Reactors", September 1997.

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9) U.S. Nuclear Regulatory Commission; " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants", NUREG 0800, June 1987.

10) U.S. Nuclear Regulatory Commission; " Assumptions Used For Evaluating The l

Potential Radiological Consequences Of A Loss Of Coolant Accident For Boiling i

Water Reactors", Regulatory Guide 1.3, June 1974.

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' 11)NRC Letter, L. Joseph Callan to The Commissioners; "Results of the Revised (NUREG-1465) Source Term Rebaselining for Operating Reactors", SECY 98-154, June 30,1998.

12)NRC Letter, John Hoyle to L. Joseph Callan; " Staff Requirements - SECY-98-158 -

Rulemaking Plan for Implementation of Revised Source Term at Operating Reactors", September 4,1998.

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Primary Containment Isolation Valves with Closure Times of at least 60-Seconds Valve Description Current Max Proposed Max Isolation Time Isolation Time (Seconds)

(Seconds)

D23-F592-A Drywell Fission Prod.

60 110 Monitor Sample D23-F591-B Drywell Fission Prod.

60 110 Monitor Sample D23-F594-A Drywell Fission Prod.

60 110 Monitor Sample Return D23-F593-B Drywell Fission Prod.

60 110 Monitor Sample Return E12-F023-A

!U i< to Head Spray 94 110 E12-F394-B 756fR to Head Spray 60 110 E12-F028A-A RHR Heat Exchanger"A" to 90 110 CTMT SPR SpargerINL E12-F037A-A RHR Heat Exchanger"A" to 74 110 CTMT Pool E12-F028B-B RHR Heat Exchanger "B" to 90 110 CTMT SPR SpargerINL l

E12-F037B-B RHR Heat Exchanger "B" to 74 110 CTMT Pool E12-F011 A-A RHR "A" Test Line to 60 110 Suppression Pool E12-F011B-B RHR "B" Test Line to 60 110 Suppression Pool l

E12-F021-B RHR "C" Test Line to 144 144 Suppression Pool E12-F012-A LPCS Test Line 144 144 I

E12-F024A-A RHR Test Return Line 144 144 L

E12-F024B-B RHR Test Return Line 144 144 E22-F023-C HPCS Test Line 75 110 E51-F076-B Steam Supply to RCIC 60 110 Turbine E51-F031-A RCIC Pump Suction 60 110 E51-F077-A RCIC Turbine Exhaust 60 110 E51-F078-B RCIC Turbine Exhaust 60 110 Vacuum Breaker G36-F106-(B)

RWCU Backwash to C/U 60 110 Phase Separator Tank G36-F101-(A)

RWCU Backwash to C/U 60 110 Phase Separator. Tank i

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Primary Containment Isolation Valves With Closure Times of at least 60-Seconds j

(continued)

Valve Description Current Max Proposed Max Isolation Time Isolation Time (Seconds)

(Seconds)

G41-F028-A FPC & CU to Upper Cont.

60 110 Pool 1

G41-F029-A Upper Cont. Pool to Fuel 60 110 Pool Drain Tank i

G41-F044-B Upper Cont. Pool to Fuel 60 110 Pool Drain Tank M71-F594-B Ctmt. Press. Inst. (Post Acc.

60 110 Sample)

M71-F595-A Ctmt. Press. Inst. (Post Acc.

60 110

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Sample)

P11-F075-(A)

Condensate Supply 60 110 P11-F130-(A)

Refueling Water Transfer 60 110 Pump Suction P11-F131-(B)

Refueling Water Transfer 60 110 Pump Suction P21-F017-A Demin. Water Supply to 60 110 Cont.

P21-F018-B Demin. Water Supply to 60 110 Cont.

P52-F105-(A)

Sen> ice Air Supply 60 110 P53-F001-(A)

Instrument Air Supply 60 110 P53-F003-A Instrument Air to ADS 60 110 P60-F009-A Suppression Pool Clean-up 60 110 Return P60-F010-B Suppression Pool Clean-up 60 110 Return P71-F150-(A)

Chilled Water Supply 60 110 P71-F148-(A)

Chilled Water Return 60 110 i

P71-F149-(B)

Chilled Water Return 60 110 I

P72-F123-B Drywell Chilled Water 60 110 Return P72-F122-A Drywell Chilled Water 60 110 Return P72-F121-A Drywell Chilled Water 60 110 Supply

Attachmmt 2 Y,

Page1 Primary Containment Isolation Valves with Closure Times not more than 60 Seconds Valve Description Current Proposed Max Max Isolation Closure Time Time (Seconds)

(Seconds)

B21-F028A Main Steam Outboard Isolation 5

NO CHANGE B21-F028B Main Steam Outboard Isolation 5

NO CHANGE B21-F028C Main Steam Outboard Isolation 5

NO CllANGE B21-F028D Main Steam Outboard Isolation 5

NO CilANGE B21-F022A Main Steam Inboard Isolation 5

NO CilANGE l

B21-F022B Main Steam Inboard Isolation 5

NO CilANGE B21-F022C Main Steam Inboard Isolation 5

NO CHANGE B21-F022D Main Steam Inboard Isolation 5

NO CHANGE E12-F008-A RHR Shutdown Cooling Outboard 40 NO CliANGE E12-F009-B RHR Shutdown Cooling Inboard 40 NO CilANGE M41-F011-A Cont. Vent and Purge Supply 4

110 M41-F012-B Cont. Vent and Purge Supply 4

110 M41-F034-B Cont. Vent and Purge Exhaust 4

110 M41-F035-A Cont. Vent and Purge Exhaust 4

110 E51-F063-B Steam Supply to RCIC Turbine 60 NO CHANGE E51-F064-A G33-F028-B Steam Supply to RCIC Turbine 60 NO CilANGE RWCU to Main Condenser 35 NO CllANGE G33-F034-A RWCU to Main Condenser 35 NO CHANGE G33-F039-A RWCU to Feedwater 35 NO CHANGE G33-F040-B RWCU to Feedwater 35 NO CllANGE G33-F001-B RWCU Pump Suction 35 NO CilANGE G33-F004-A RWCU Pump Suction 35 NO CHANGE G33-F252-B RWCU Pump Suction 35 NO CHANGE G33-F053-B RWCU Pump Discharge 35 NO CilANGE G33-F054-A RWCU Pump Discharge 35 NO CilANGE B21-F067A-A Main Steam Line Drain 9

NO CllANGE B21-F067B-A Main Steam Line Drain 9

NO CllANGE B21-F067C-A Main Steam Line Drain 9

NO CllANGE B21-F067D-A Main Steam Line Drain 9

NO CilANGE B21-F019-A Main Steam Line Drain 20 NO CilANGE B21-F016-B Main Steam Line Drain 20 NO CHANGE

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Primary Containment Isolatlan Valves with Closure Times not mo.c than 60 Seconds l

Valve Description Current Proposed Max Max Isolation Isolation Time Time (Seconds)

(Seconds)

E61-F009-A Containment Purge Supply 4

110 E61-F010-B Containment Purge Supply 4

110 E61-F056-B Purge Filter Train Isolation 4

110 E61-F057-A Purge Filter Train isolation 4

110 P45-F061-B Cont. Floor Drain Discharge 7

110 i

P45-F062-A Cont. Floor Drain Discharge 7

110 P45-F067-B Cont. Floor Drain Discharge 7

110 P45-F068-A Cont. Floor Drain Discharge 7

110 P45-F098-B Chemical Waste Sump Discharge 8

110 P45-F099-A Chemical Waste Sump Discharge 8

110 i

P45-F273-A Aux. Bldg. E/F Drain to Suppression 32 110 Pool P45-F274-B Aux. Bldg. E/F Drain to Suppression 32 110 Pool l

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Attcchment 43 Page1 Evaluation of Significant Hazards Considerations And Environmental Impact SIGNIFICANT HAZARDS EVALUATION GGNS staff has evaluated the proposed change to incorporate a delay in the post-accident l

fission product release into its licensing basis. This change recognizes one of the revised source term insights discussed in NUREG 1465. This change in the licensing basis will l

provide the basis for revising the Technical Requirements Manual to increase Primary Containment Isolation Valve (PCIV) maximum isolation times. These changes have been evaluated using the standards in 10CFR50.92 and it is concluded that they do not involve any significant hazards considerations. Specifically, the proposed change will not:

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1) Involve a significant increase in the probability or consequences of an accident previously l

evaluated, The proposed change takes credit for a new source term insight that recognizes that the fission product release from a fuel assembly is not instantaneous with a design basis accident. Implementation of this change into the licensing basis will be used tojustify an f

increase in the maximum allowable PCIV isolation times. These changes do not affect l

the precursors for any accident or transient evaluated in Chapter 15 of the GGNS UFSAR. Therefore, there is no increase in the probability of any accident previously evaluated.

l A plant specific radiological analysis has been performed to evaluate the effect on the l

dose consequences of extending the maximum allowable closure time. This evaluation considered the initial two-minute period of the accident during which, according to new source term insights developed in NUREG-1465 and in a BWROG report, fission product releases are not expected to occur. Releases from the break and from containment during l

this period consist of coolant radioactivity only. The total release during this period was j

found to result in an offsite dose ofless than 0.60 rem. This dose represents only a small fraction of the LOCA dose evaluated in the UFSAR. As this submittal is for a limited scope application of the NUREG-1465 insights (in this case, timing and duration of the coolant activity phase) and addresses only the first 121 seconds of the accident scenario, the total long-term dose determined using the TID-14844 assumptions is not changed by this submittal. In reality, the other insights offered in the NUREG would be expected to result in an overall dose reduction. In any event, the dose consequences of the proposed change do not result in an increase in the consequences of any accident previously evaluated.

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2) Create the possibility of a new or different kind of accident from any accident previously evaluated; The primary containment isolation system is designed to prevent, as much as practicable, the unfiltered release ofradioactive material to the environs following an accident. As such, the system is relied upon for accident dose consequence mitigation. Neither the revision of the licensing basis to recognize that fission product releases are not instantaneous as is assumed in the current analysis, nor the extension of the valve closure times affects the ability of the valves to perform their accident mitigation function. It is also noted that the increased closure time allowables will only be applied to valves which do not have an alternate constraining performance requirement for closure time; the safety functions of other supported components and systems are not affected. Thus, the proposed change does not create the potential for a new or different kind of accident.

3) Involve a significant reduction in a margin of safety.

The proposed change revises the bases for the offsite dose calculation to credit, in the initial 2 minutes of the accident scenario, the fact that there is no fuel failure expected during this time. That is, for the first two minutes of the event, only coolant activity is released. The other assumptions, bases and methodologies for offsite dose calculations used to evaluate the long-term offsite dose consequences of accidents described in FSAR Chapter 15 are not affected by this change. The margin between calculated dose consequences described in the FSAR and regulatory limits is not reduced.

A recent GGNS analysis of the LOCA scenario considering the only release in the first 121 seconds is from the reactor coolant resulted in an EAB dose ofless than I rem thyroid during this period. The total dose for the 0- to 2-hour period is not expected to increase due to the delay in the fission product release; the total amount of radioactivity released will remain the same. Both the recently evaluated 2-minute dose and the 24.9 rem in two hours as presented in the UFSAR are insignificant in comparison to the 300 rem acceptance limit for this scenario. The GGNS SER acknowledges the conservatism of the old analysis methodology. An independent analysis done by the staff during their evaluation of the GGNS FSAR estimated doses could decrease about 95% if the fission product release were to be delayed by 2 minutes.

The bases for PCIV closure times described in the Technical Specifications remain unchanged. The inconsistency between the assumption ofimmediate containment isolation in the dose analysis and allowable isolation valve closure times of one to two minutes is eliminated by this change. Plant specific analysis has shown that the expected dose resulting from the PCIVs remaining open during this period is insignificant.

Actual safety benefits are expected to result from valve performance and reliability improvements, elimination of unnecessary reports and system performance improvements such as minimization of water hammer events. Therefore, the increase in maximum l

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' 3 Page1 isolation time for certain PCIVs proposed in this submittal will not result in a significant reduction in the margin ofsafety.

l ENVIRONMENTAL IMPACT CONSIDERATION The proposed change does not involve a significant hazards consideration, a significant change in the types of, or a significant increase in the amounts of, any effluents that may be released offsite, or a significant increase in an individual or in the cumulative occupational radiation exposure. Therefore, the proposed change meets the eligibility criteria far categorical exclusion set forth in 10CFR51.22 (c.)(9). Accordingly, pursuant to 10CFR51.22 (b), an environmental assessment of the proposed change is not required.

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. Excerpts and Markup of Section 15.6.5.5.2 to reflect the Proposed Licensing Basis Change i

EXISTING:

15.6.5.5.2 Containment Activity Inventory The activity released from the severely damaged core enters the drywell at accident time zero. Transfer from the drywell to the containment is either through the suppression pool, where a decontamination factor of 10 is taken, or through drywell leakage, which bypasses the suppression pool. This bypass flow is assumed to be equally divided between containment regions 1,3, and 4 defined below. The flowrates for each of these drywell release pathways is based on the pressure differential between the drywell and containment (see Section 6.2). Suppression pool scrubbing (with a DF of 10) is assumed to remain effective as long as there is flow from the drywell into the suppression pool.

PROPOSED:

The activity released from the severely damaged core enters the drywell at 121 seconds after the accident. This timing assumption recognizes conclusions derived from source term studies as described in NUREG 1465. Transfer from the drywell to the containment is either through the suppression pool, where a decontamination factor of 10 is taken, or through drywell leakage, which bypasses the suppression pool. This bypass flow is assumed to be equally divided between containment regions 1,3, and 4 defined below. The flowrates for each of these drywell ielease pathways is based on the pressure differential between the drywell and containment (see Section 6.2). Suppression pool scrubbing (with a DF of 10) is assumed to remain effective as long as there is flow from the drywell into the suppression pool.

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