ML20195K259

Summary of 981001 Meeting with NEI in Rockville,Md Re Status of Staff Efforts to Allow Voluntary Implementation of Revised Source Term at Operating Reactors.List of Attendees Encl Also

ML20195K259
Person / Time
Issue date:	11/20/1998
From:	Stewart Magruder NRC (Affiliation Not Assigned)
To:	Essig T NRC (Affiliation Not Assigned)
References
FRN-64FR12117, PROJECT-689 AG12-1-045, AG12-1-45, NUDOCS 9811250217
Download: ML20195K259 (45)

Text

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WASHINGTON, D.c. 2006M001

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November 20, 1998 MEMORANDUM TO: Thomas H. Essig, Acting Chief Generic issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation FROM:

Stewart L. Magruder, Project Manager Generic issues and Environmental Projects Branch Division of Reactor Program Management Office of Nuclear Reactor Regulation

SUMMARY

OF OCTOBER 1,1998, MEETING WITH THE NUCLEAR ENERGY INSTITUTE (NEI) REGARDING THE STATUS OF STAFF EFFORTS TO ALLOW VOLUNTARY IMPLEMENTATION OF THE REVISED SOURCE TERM AT OPERATING REACTORS On October 1,1998, reprecentatives of the Nuclear Energy Institute (NEI) met with representatives of the Nuclear Regulatory Commission (NRC) at the NRC's offices in Rockville, Maryland. Attachment 1 provides a list of meeting attendees. The purpose of the meeting was to provide an opportunity for the NRC staff to update the members of the NEl task force on the implementation of the revised (NUREG-1465) source term at operating reactors.

The staff begh mth a presentation of the results of a rebaselining study on the impact of implementing r,e revised L,0urce term at operating reactors. The staff stated that the study did not identify any issues that would prevent implementation and that the rebaselining activities provided a technical basis for rulemaking and the associated regulatory guides (RGs). A copy of this presentation is included as Attachment 2.

The staff next answered several questions about implementation issues with the revised source term:

Q1. Since lodine is now in particulate form, can licensees take credit for HEPA filters removing some percentage of it?

A. Yes. The staff will use a value of 99% for the pilot plants and will review this percentage during the development of a regulatory guide.

Q2. Will the length of time that equipment is assumed to function after an accident

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(qualified life for environmental qualification purposes) be changed as a result of the revised source term?

A. It may, however, the answer depends on each plant's licensing basis.

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-2 November 20, 1998 1

Q3. What is the availablity of the RADTRAD computer code?

A. It is still being refined. When it is in naal form it will be available from Oak Ridge National Laboratory.

Q4. What is the link between doses calculated for accident analyses and those calculated for environmental qualification purposes?

i A. This issue is still being discussed within the staff.

The staff next made a presentation on the rulemaking plan. An accelerated schedule was shown to the group and is included as Attachment 3. The staff noted that a draft RG and a draft standard review plan (SRP) will be published with the final rule in September 1999, and that there will be many opportunities to discuss the RG and SRP before then.

A preliminary draft of the rule language was also shown to the group and is included as. The staff briefly discussed the Seotember 4,1998, staff requirements memorandum (SRM) on SECY-98-158, "Rulemaking Plan for Implementation of Revised Source Term at Operating Reactors." It was noted that the SRM allows plants, other than the pilot plants, to start applying for exemptions to allow them to implement the revised source term. The group also discussed the difficulty of selectively implementing the revised source term.

The staff next made a brief presentation on the status of the review of pilot plant submittals.

The presentation materialis included as Attachment 5.

Finally, e representative from Entergy gave a presentation on the pilot activities underway at the Grand Gulf Nuclear Station. This materialis included as Attachment 6. The Entergy representative noted that just implementing a new timing methodology veould give them relief on containment isolation valve testing and that this was tied to the approval of d boiling water reactor owners group report. The group agreed that another meeting on reviserl source term implementation issues should be scheduled for early 1999.

Project No. 689 Attachments: As stated cc w/att: See next page DISTRIBUTION: See attached page OFFICE PM:PGEB PERB SC:PGEB NAME SMagruder:# SlaVie f kk 11/ 4 /98 11/fbO8 DATE 11/ F1/98 1

1

l T. Essig

-2 November 20, 1998 i

Q3. What is the availablity of the RADTRAD computer code?

A. It is still being refined. When it is in final form it will be available from Oak Ridge National Laboratory.

Q4. What is the link between doses calculated for accident analyses and those calculated for environmental qualification purposes?-

i A. This issue is still being discussed within the staff.

I The staff next made a presentation on the rulemaking plan. An accelerated schedule was l

shown to the group and is included as Attachment 3. The staff noted that a draft RG and a draft standard review plan (SRP) will be published with the final rule in September 1999, and that there will be many opportunities to discuss the RG and SRP before then.

A preliminary draft of the rule languaoe was also shown to the group and is included as. The staff briefly discussed the September 4,1998, staff requirements I

memorandum (SRM) on SECY-98-158, "Rulemaking Plan for implementation of Revised Source Term at Operating Reactors." It was noted that the SRM allows plants, other than the pilot plants, to start applying for exemptions to allow them to implement the revised source term. The group also discussed the difficulty of selectively implementing the revised source term.

The staff next made a brief presentation on the ctatus of the review of pilot plant submittals.

l The presentation material is included as Attachment 5.

Finally, a representative from Entergy gave a presentation on the pilot activRies underway at the l-Grand Gulf Nuclear Station. This materialis included as Attachment 6. The Entergy l-representative noted that just implementing a new timing methodology would give them relief l'

on containment isolation valve testing and that this was tied to the approval of a boiling water reactor owners group report. The group agreed that another meeting on revised source term implementation issues should be scheduled for early 1999.

Project No. 689 l

Attachments: As stated cc w/att: See next page L

l NEl/NRC MEETING ON REVISED SOURCE TERM i

10/1/98 List of Attendees 1

l Name Oraanization l

'Kurt Cozens NEl l

Marvin Smith VEPCo Douglas Gilliatt VEPCo John Duffy PSE&G

' John Nagle PSE&G Sreela Ferguson Stone & Webster

' Tom Mscisz PECO Energy Tom Milton Southern Nuclear

' Jason Chao -

EPRI David Leaver Polestar i

Charles Jackson Coned Ken Jha Bechte! -

'Greg Broadbent Entergy.

i Mike Withrow Entergy

- Al Widmer Centerior Energy

- Bradley Ferrell '

Centerior Energy

_ i Christopher Smith NUSIS j,

Mike Radvansky GPU Nuclear

- Theresa Sutter.

Bechtel Lynn Connor N/A

- James Grover Westinghouse l

Douglas Pickett NRC/NRR l

Barry Zaleman -

NRC/NRR Steve LaVie NRC/NRR.-

Mark Blumberg.

NRC/NRR

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NRC/NRR

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Jay Lee' k

Rich Emch NRC/NRR -

i.

Stu Magruder NRC/NRR L-Jocelyn Mitchell NRC/EDO Charles Ader -

NRC/RES Jason Schaperow NRC/RES Charles Tinkler '

NRC/RES Chester Gingrich-NRC/RES 4

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United States Nuclear Regulatory Commission g,.

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l Results of Revised (NUREG-1465;) Source Term Rebaselining for Operating Reactors r

Meeting with the Nuclear Energy institute i

f l

Jason H. Schaperow Office of Nuclear Regulatory Research i

f October 1,1998 t

i t

i f

a-i Rebaselining Objective was to develop a better understanding of the impacts of

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implementing.the revised source term for operating reactors t

Effect on calculation of individual offsite and control room dose

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Effect on calculation of dose for equipment-qualification Effect of potential plant modifications, including severe accident risk l

impacts Rebaselining was an assessment of the likely dominant issues as revealed by analysis of representative plants and served as " test bed" for i

developing regulatory criteria Rebaselining results provided in SECY-98-154 i

l 2

COMPARISON OF TID-14844 AND NUREG-1465 SOURCE TERMS (PWR) i i

TID-14844 NUREG-1465 i

Instantaneous Release Release over 1.8 Hrs 100% Noble Gases 100% Noble Gases 50% lodines (with 50% Plateout) 40% lodines 1% Solids 30% Cesium 5% Tellunum j

2% Barium j

.02%

.2% Others Indine -

91 % inorganic vapor 4.85% inorganic vapor i

4% organic vapor

.15% organic vapor j

5% aerosol 95% aerosol Solids Solids normally Solids treated I

ignored for offsite dose as aerosol calculation i

3 f

. -.. =.

1 h

- a Rebaselining Tasks Phasei DBA Dose Calculations (SER modelling)

Phase 11 DBA Dose Calculations (FSAR modelling)

Phase ill DBA Dose Calculations (" updated" fission product removal models)

Thermal-Hydraulic Considerations Assessment of Margins Sump pH and lodine Revolatilization i

Phase IV DBA Dose Calculations (plant changes)

Risk Impacts of Plant Changes I

I t

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i Phasei I

DBA Dose Calculations (SER modelling) l Objectives were to determine the effect on individual dose of substituting NUREG-1465 for TID-14844 source term and the effect

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of the new dose acceptance criteria using the SER modelling.

L Analyses performed for Grand Gulf and Surry.

Accidents analyzed include LOCA, Fuel Handling, MSLB, Rod Drop Calculations performed for EAB, LPZ and control room f

i 5

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Phase 11 i

l DBA Dose Calculations (FSAR modelling)

Objectives Determine the effect on individual dose of substituting NUREG-1465 for TID-14844 source term and the effect of the new dose l

acceptance criteria utilizing the FSAR modelling.

j l

Determine the effect on equipment qualification dose of substituting NUREG-1465 for TID-14844 source term.

Analyses performed for Grand Gulf and Surry Additional calculations performed for Zion, typical of a large, dry containment.

6 l

t

9 Grand Gulf LOCA - Doses frem) Phase ll l

Source-EAB EAB EAB TEDE LPZ LPZ LPZ l

Term Thyroid Whole Thyroid Whole TEDE Body Body

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TID 23.1 9.5 NA 40.1 7.57 NA 11.5(1.0h) l NUREG 10.6 1.5 2.0 19.5 4.05 4.73 L

22.6(1.5h)*

5.9(2.3h) 6.8(2.2h)

• Worst two hours dose and start of worst two hoisrs Findings j

i i

Substantial difference between 1st 2 hr. dose and worst 2.hr. dose.

LPZ thyroid dose is reduced due to lower organic iodine and smaller (iodine) source term for ECCS leakage.

Major contribution to TEDE from noble gases.

7 I

- - -- _ j

i 1

Phase ill i

Objective:

Assessment of detailed analysis assumptions and technical r

issues associated with implementation of the revised source 4

term and evaluation of conservatism in regulatory analysis.

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i Tasks l

q 1.

Perform dose calculations for Surry, Grand Gulf, and Zion using l

NUREG-1465 and updated removal mechanism modeling: SRP I

models and models described in NUREG/CR-5966 (sprays),

NUREG/CR-6153 (suppression pools).

2.

Perform best estimate analysis using MELCOR for Surry, Grand Gulf, and Zion to assess margins in DBA LOCA analysis i

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i Assumptions EAB LPZ Thyroid Whole TEDE Thyroid Whole TEDE Body Body.

SER 232 3.09 NA 20.3

.270 NA (TID-14844) l e

FSAR 225 3.36 NA 12

.15 NA (TID-14844) f i

UPDATED MODELS 76

.46 3.55 4.41

.021

.19 l

(NUREG-1465)

MELCOR 1.55

.006

.055 1.16

.001

.037 (NUREG-1465) l

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Comparison of Phases I, il, and ill Surry LOCA Doses (rem) i i

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EQ Dose Analysis-EQ doses calculated for containment atmosphere and sump using TlO and NUREG-1465 source terms Gamma and beta doses in containment atmosphere Similar doses between TID and NUREG-1465 source terms, because the dose is from noble gases and iodine Gamma dose for equipment exposed to sump water Higher at later times for the NUREG-1465 source term, because of the large amount of cesium in the NUREG-1465 source term.

TID-14844 includes 1% of the core inventory of cesium, NUREG-1465 includes 30% of the core inventory of cesium.

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Surry Gamma Dose At Containment Center l@ i

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1 07 !

2 TID-14844 i

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1 06 3 i

1 05 i::

l@ 1 NNUREG-1465 1 03 10-1 100 1 01 1 02 1 03 1 04 Time (hr)

Surry Beta Dose At Containment Center 17 i:

108 h TID-14844

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107 i 3

S.1 05 1 i

-N U REG-1465 l@ i 10' i

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Surry Sump Gamma Dose 1 08 y t-ip 107$

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1 06

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No Cs 105h:

NUREG-1465 10 -

102-1 00 1 01 102 1 03 1 04 Time (hr)

Figure 3

k Phase IV DBA Dose Calculations (Plant Changes) i Performed offsite and control room dose calculations for LOCA for l

Surry, Zion, and Grand Gulf

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Used FSAR plant models and updated removal mechanism models.

i Changes considered to 1) MSIV leakage control system, 2) containment recirculation filters,3) charcoal filters,4) containment leak rate, 5) spray startup time,6) enclosure building drawdown time, 7) changing from subatmospheric to atmospheric containment 9

Compared results to earlier results without plant changes and to f

current and proposed dose limits.

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Elimination of MSIV Leakaae Control System (LCS) and increase in Allowable MSIV Leakage Perry pilot plant proposed eliminating the MSIV leakage control system and increasing the MSIV leak rate from 100 to 250 scfh.

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Assessed changes using Grand Gulf plant model.

f Removal of the MSiV leakage control system resulted in doses less than the dose !!mits.

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Some increase in the allowable MSIV leak rate was possible without exceeding the dose limits if credit for deposition in the main steam line is permitted. Not as much increase as proposed by Perry was acceptable.

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I Elimination of MSIV Leakage Control System and increase in Allowable MSIV Leakaae Ccce EAB LPZ Control Room Thyroid Whole TEDE Thyroid.

Whole TEDE Thyroid Whole TEDE

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-Body Body Body Phnse ill 24.4(1.6h) 5.90(2.3h) 6.98(2.2h) 19.9 4.06 4.76 4.06

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.54 No LCS 226(.3h) 5.72(2.5h) 14.5(.5h) 54.4 4.22 6.82 9.34

.40

.85 No LCS, 2169(.3h)

.14.2(.5h) 119(.4h) 492 6.72 30.4 81.2

.57 4.49 2.2%/ day

. No LCS, 852(.3h) 10(.5h) 50(.4h) 196 5.73 15.0 32.6

.51 2.05 2.2%/ day, steam line deposition Note: 2.2%/ day corresponds' roughly to 250 scfh Based on Grand Gulf plant model.

t 15 l

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.l Phase IV 1

Risk Imparts from implementation of the Revised Source Term i

i Obientire The objective of the study was to evaluate the change in plant risk from modifying ESFs or their operation.

i Approach t

i Modify PRAs to model potential change in component performance for i

the various features including: 1) Containment Leak Rate, 2) Containment Spray Operation,3) Reactor Building Drawdown Time,4) l Subatmospheric Containment at Atmospheric Pressure, and 5) Filtration Systems.

l The NUREG-1150 Plants (Peach Bottom, Grand Gulf, Surry, Sequoyah, and Zion) and LaSalle were evaluated.

t 16 I

I

- - ]

L Results - Change in Plant Risk

.)

l 1)

Containment Leak Rate-+ Small (risk dominant sequences involve l

containment failure or containment bypass).

2)

Containment Spray Operation + Small (delay of sprays until fission product release has-no effect on risk and preserves stored water inventory) 3)

Reactor Building Drawdown Time + Small (effect for sequences that do not fail containment < 1% of total plant risk) 4)

Subatmospheric Containment at Atmospheric Pressure + Small (major contributors to risk are accident sequences that involve containment bypass).

i 4

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5) Filtration Systems + Small (filters generally not credited for risk t

dominant sequences).

j i

l 17 l

I

Conclusions Offsite and Control Room DBA Doser.

i Impact of NUREG-1465 vs TfD-14844 is to generally produce lower calculated doses Extent of the reduction is influenced by several factors Influence of safety features which are timing sensitive (e.g.,

SGTS, subatmospheric design)

Analysis assumptions used in SER and FSAR calculations Use of updated dose conversion factors will, by itself, produce lower calculated doses Many of the types of plant changes being contemplated could be made and offsite and control room DBA doses would remain within acceptance limits.

18

Conclusions (2)

Equipment Qualification Doses a

Similar doses for equipment exposed to containment atmosphere.

l Higher doses later in time for equipment exposed to sump water, due to higher cesium inventory in NUREG-1465 source term.

Significance of higher sump water doses will be considered in the pilot plant reviews.

r I

L 19

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Conclusions (3)

Margin and Risk Analysis performed with MELCOR severe accident code indicated that the offsite DBA doses still have substantial margin (a factor of 2 or greater) even though the dose may be well below the earlier TID analysis.

Potential plant changes being contemplated with the NUREG-1465 source term are not likely to have substantial risk impacts, because most of the systems being changed are not involved in risk significant sequences.

Implementation of NUREG-1465 Source Term The staff did not identify any issues that would prevent implementation of the revised source term at operating reactors.

The rebaselining activities have provided a technical basis for rulemaking and the associated regadatory guides.

20

i Slupporting Documents t

i in response to requests for documents describing details of the rebaselining analyses, the following reports will be placed in the NRC Public Document Room:

i i

NRC report AEB-98-01, " Detailed Modeling and Analysis of Phase ill

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of the Revised (NUREG-1465) Source Term Rebaselining fc,t

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Operating Reactors, Issued September 23,1998 j

i NRC report AEB-98-02, " Detailed Modeling and Analysis of the Zion Plant for Phase ll of the Revised (NUREG-1465) Source Term Rebaselining for Operating Reactors," To Be issued in October 1998 f,

SNL report entitled " Evaluation of Radiological Consequences of Design Basis Accidents at Operating Reactors Using the Revised i

Source Term," issued September 28,1998 i

i i

21 i

J Supporting Documents (2)

ORNL report ORNL/NRC/LTR-97/21, "MELCOR DBA LOCA Calculations," To Be issued in October 1998 ORNL report ORNL/M-6544, " lodine Revolatilization in a Grand Gulf LOCA," To Be issued in October 1998 i

NUREG/CPc6418, " Risk Importance of Containment and Related ESF System Performance Requirements," To Be issued in October 1998 NUREG/CR-6604, "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," Issued April 1998 NUREG/CR-6210, " Computer Codes for Evaluation of Control Room Habitability (HABIT)," Issued June 1996 E

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- - - - - - - - - ~ - ~ ~ - - - - - ~ ~ ~ ^ ~ ~ ~

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j With Original Buildup Factors 1

108 Sandia/

107 --! TI D-14844 H

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NUREG-1465 m

No Cs 8

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E NUREG-1465 103

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103 104 Time (hr)

With Revised Buildup Factors 108

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NUREG-1465 m

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Figure 5-14.Surry Sump Doses.

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101 100 101 102 103 104 Tirne (hr)

With Revised Bui/ dup Factors 108.

107,

g Sandia/

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105 -

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104 -

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Figure 5-17. Grand Gulf Sump Deses.

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i ID Task Name l Duraticl Start l

Finish 8/30 l 9/20 l 1"11 l 11/1 l 11/22 l 12/13 l 10 l 1/24 l 2/14 l 3/7 l 3/26 l 4/18 l 5ft '

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1 Proposed Rule Package 106d 8/31/98 1/25/99 2

Develop rute package 25d 8/31/98 10/2/98 3

Branch /Drvision/NRR concurrence 5d 10/5/98

• #9/98 4

Office concurrence 15d 10/12/98 10/30/98 5

ACRS Review 15d 10/19/98 11/6/98 5

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6 ACRS Meeting Od 11/5/98 11/5/98 f3 7

CRGR Concurrence 20d 11/5/98 12/2/98

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8 CRGR Meeting Od 11/27/98 11/27/98 g7 9

EDO Concurrence 5d 12/3/98 12iV98 l

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10 Submit proposed rule to Commissi Od 12/15/98 12/15/98 f.

11 Commission Review 20d 12/15/98 1/11/99 12 Resolve comments; FR publish 10d 1/12/99 1/25/99 j

13 Final Rule; Draft Guide-1081; Draft SRP 164d 1/26.99 9/10/99 j

14 Public Comment Period 60d 1/26/99 4/19/99

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g 15 Resolve Public Comments 15d 4/20/99 5/10/99 O

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16 Develop Regulatory Guide 75d 1/26/99 5/10/99 j

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!!f 17 Prepare final rule package 15d 5/11/99 5/31/99

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18 Branch / Division /NRR Sd 6/1/99 6/7/99 5

l 19 Office Concurrence 15d 6/8/99 6/28/99 5

20 ACRS Review 15d 6/14/99 7/2/99 I

21 CRGR Concurrence 15d 7/5/99 7/23/99 5

22 EDO Concurrence 5d 7/26/99 7/30/99 5

23 Final Rule, draft guide to Commissi Od 7/30/99 7/30/99 5

REVISED SOURCE TERM PROPOSED & FINAL RULE Progress m Rolled Up Task lj; i[j DG-1081 & SRP 15.0.1 Milestone Rolled Up Milestone Q au

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ID l Task m:ne lDuratic l Start l

Finish 5/30 l 6/20 l 7/11 l 01 l 8/22 l 9/12 l 10/3 l 10/24 l 11/14 l 12!5 l 12/26 l 1/16 l 2/6 "

1 Prop. sed Rule Package 106d 8/31/98 1/25/99 2

Develop rule package 25d 8/31i98 10/2/98 3

Branch / Division /NRR concurrence 5d 10/5/98 10/9/98 4

Office concurrence 15d 10/12/98 10/30/98 5

' ACRS Review 15d 10/19/98 11/6/98 6

ACRS Meeting Od 11/5/98 11/5/98 7

CRGR Concurrence 20d 11/5/98 12/2/98 8

CRGR Meeting Od 11/27/98 11/27/98 9

EDO Concurrence 5d 12/3/98 12/9/98 10 Submit proposed ru!e to Commissi Od 12/15/98 12115/98 11 Commission Review 20d 12/15/98 1/11/99 12 Resolve comments: FR publish 10d 1/12/99 1/25/99 13 Final Rule: Draft Guide-1081; Draft SRP 164d 1/26/99 9/10.99 14 Public Comment Penod 60d 1/26/99 4/19/99 15 Resolve Public Comments 15d 4/20/99 5/10/99 16 Develop Regulatory Guide 75d 1!26/99 5/10/99 J

17 Prepare final rule package 15d 5/11/99 931/99 1

18 Branch / Division /NRR Sd 6fi/99 6/7/99 19 Office Concurren 15d 6/8/99 6/28/99 20 ACRS Review 15d 6/14/99 7/2/99 21 CRGR Concurrence 15d 7/5/99 7/23/99 gp 22 EDO Concurrence 5d 7/26/99 7/30/99 23 Final Rule, draft guide to Commissi Od 7/30/99 7/30/99 7/30 f REVISED SOURCE TERM PROPOSED & FINAL RULE Progress

- Rolled Up Task lq y-l DG-1081 & SRP 15.0.1 Milestone Roi?ed Up Milestone Q Fage 2

d ID ITask Name I Duratic l Start l

Finish 5/30 l 6/20 l 7/11 l 8/1 l 8/22 l 9/12 l 10/3 l 10/24 l 11/14 l 12/5 l 12/26 l 1116 l 2/6

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qp 24 Commission Review 20d 8/2/99 8/27/99 25 Resolve mmments; FR publish 10d 8/30/99 9/10/99 26 Final Guide; Final SRP 15.0.1 95d S/13/99 1/24/00 y.

27 Public Comment Period 33d 9/13/99 10/27/99 V////////////

28 Resolve Public Comments 12d 10/28/99 11/12/99

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-j 29 Prepare final Guide & SRP; FR notice 10d 11/15/99 11/26 S9 1

^N 30 Branch / Division /NRR Sd 11/2999 12/3/99 31 Office Concurrence 10d 12/6/99 12/17/99

'C 32 ACRS Review 10d 12/13/99 12/24/99

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'"Y 33 CRGR Concurrence 15d 12/27!99 1/14/00

....b 34 EDO Concurren 5d 1/17/00 1/21/00 35 Final guide & SRP to Commission Od 1/24/00 1/24/00

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w REVISED SOURCE TERM PROPOSED & FINAL RULE Progress RoEed Up Task li; jij DG-1081 & SRP 15.0.1 Milestone Roned Up Milestone Q Page 3

PRELIMINARY DRAFT PRELIMINARY DRAFT PRELIMINARY DRAFT PART 21 -- REPORTING OF DEFECTS AND NONCOMPLIANCE

1. The authority citation for Part 21 continues to read as follows:

Authority: Lec.161,68 Stat. 948, as amended, sec. 234,83 Stat. 444, as amended, sec.1701, 106 Stat. 2951,2953 (42 U.S.C. 2201,2282,2297f); secs. 201, as amended,206,88 Stat.1242, as amended, 1246 (42 U.S.C. 5841,5846).

Section 21.2 also issued under secs. 135,141, Pub. L. 97 - 425, 96 Stat. 2232,2241 (42 d.S.C.

10155,10161).

2. Section 21.3 is amended by revising paragraph (1)(i)(C) of the definition of Basic Component to read as follows:

$21.3. Definitions.

As used in this part:

Basic component.

(1)(i)

(C) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in @50.34(a)(1),

@50.67(b)(2), or 100.11 of this chapter, as applicable.

PAR 7 50 - DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES L The authority citation for Par 150 continues to read as follows:

Authority: Secs. 102,103,104,105,161,182,183,186,189,68 Stat. 9.36,937,938,948,953, 954,955,956, as amended, sec. 234,83 Stat. 444, as amended (42 U.S.C. 2132,2133,2134,2135, 2201,2232,2233,2236,2239,2282); secs. 201, as amended,202,206,88 Stat.1242, as amended, 12M 1246 (42 U.S.C. 5841,5842,5846).

Section 50.7 also issued under Pub. L. 9509601, sec.10,92 Stat. 2951 (42 U.S.C. 5851).

Section 50.10 also issued under secs. 101,185,68 Stat. 955 as amended (42 U.S.C. 2131,2235),

sec.102, Pub. L. 9109190,83 Stat. 853 (42 U.S.C. 4332). Sections 50.13,50.54(dd), and 50.103 also issued under sec.108,68 Stat. 939, as amended (42 U.S.C. 2138). Sections 50.23,50.35, 50.55, and 50.56 also issued under sec.185,68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a,50.55a and Appendix Q also issued under sec.102, Pub. L. 9109190,83 Stat. 853 (42 U.S.C. 4332).

Sections 50.34 and 50.54 also issued under sec. 204,88 Stat.1245 (42 U.S.C. 5844). Sections 50.58,50.91, and 50.92 also issued under Pub. L. 9709415,96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec.122,68 Stat. 939 (42 U.S.C. 2152). Sections 50.800950.81 also issued under sec.184,68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued under sec.187, 68 Stat. 955 (42 U.S.C 2237).

4. Section 50.2 is amended by revising paragraph (1)(iii) of the definition of Basic Component and by adding in alphabetical order the definition for source term to read as follows:

$50.2 Definitions.

As used in this part, Basic component (1)

(iii) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in $50.34(a)(1), @50.67(b)(2), or @100.11 i

I of this chapter, as applicable.

l

PRELIMINARY DRAFT PRELIMINARY DRAFT PRELIMINARY DRAFT Source term refers to the magnitude and mix of radionuclides released from the reactor core, their physical and chemical form, and the timing of their release.

5. Section 50.34 is amended by revising paragraphs (f)(vii), (f)(viii), (i)(xxvi), and (f)(xxviii), and adding new footnote 11 to read as follows:

s50,34 Contents of applications; technicalinformation (f)

(2)

(vii) Perform radiation and shielding design reviews of spaces around systems that may, as a result of an accident, cor>tain accident source term"U radioactive materials, and design as necessary to permit adequate access to important areas and to protect safety equipment from the radiation environment. (ll.B.2)

(viii) Provide a capability to promptly obtain and analyze samples from the reactor coolant system and containment that may contain accident source term"U radioactive materials without radiation er.posures to any individual exceeding 5 rems to the whole body or 50 rems to the extremities. Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree of core damage (e.g., noble gases, radiciodines and cesiums, and nonvolatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chioride, and boron ccncentrations. (ll.B.3)

(xxvi) Provide for leakage control and detection in the design of systems outside containment that contain (or might contain) accident source term"" radioactive materials following an accident. Applicants shall submit a leakage control program, including an initial test program, a schedule for re-testing these systems, and the actions to be taken for minimizing leakage from such systems. The goal is to minimize potential exposures to workers and public, and to provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency. (Ill.D.1.1)

(xxviii) Evaluate potential pathways for radioactivity and radiation that may lead to control room habitability problems under accident conditions resulting in an accident source term"" release, and make necessary design provisions to preclude such problems. (Ill.D.3.4)

" The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of site analysis or postulated from consideratioris of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.

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6. Section 50.49 is amended by revising paragraph (b)(1)(i)(C) to read as follows:

i

$50.49 Environmental qualification of electric equipment important to safety for i

nuclear power plants.

(b)

(1)

(i)

(C) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in S50.34(a)(1),

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v PRELIMINARY DRAFT PRELIMINARY DRAFT PRELIMINARY DRAFT 650;67(b)(2), or Q100.11 of this chapter, as applicable.

7. Section 50.65 is amended by revising paragraph (b)(1) to read as follows:

@50.65 Requirements for monitoring the effectiveness of maintenance at nuclear power plants.

(b)

(1) Safety-related structures, systems and components that are relied upon to remain functional during and following design basis events to ensure the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a l

safe shutdown condition, or the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the guidelines in

@50.34(a)(1), @50.67(b)(2), or @100.11 of this chapter, as applicable.

8. Part 50 is amended to add f50.67, a new section, to read as follows:

l 550.67 Accident source term l

(a) Applicability. The requirements of this section apply to all holders of operating licenses issued prior to January 10,1997, who seek to revise the current accident source term used in their design basis radiological analyses.

(b) Requirements: (1) A licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under @50.90. The application shall contain an evaluation of the W previously analyzed in the safety consequences of applicable design basis accidents analysis report.

- (2) The Commission may issue the amendment only if the applicant's analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the onset of the postulated fission product release, would not receive l

a radiation dose in excess of 25 remm total effective dose equivalent (TEDE).

l (ii) An individuallocated at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product j

release (during the entire period of its passage), would not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE).

(iii) Adequate radiation protection is provided to permit access and occupancy of the i

control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.

l The dose criteria of Part 50 Appendix A GDC-19 do not apply.

l

' The fission product release assumed for these calculations should be based upon a major accident, l

hypothestzed for purposes of design analyses or postulated from considerations of possible accidental l

events, that would result in potential hazards not exceeded by those from any accident considered credible.

i Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent l

release of appreciable quantities of fission products.

2 The total effective dose equivalent (TEDE) of 25 rem referred to above is specified for use with revised source terms since it utilizes a risk-consistent methodology to assess the radiological impact of all relevant nuclides upon all body organs. The latent cancer risk of a radiation dose of 25 rem TEDE is consistent with the latent cancer risk associated with exposures of 25 rem to the whole body and 300 rem to the thyrold. Risk of latent cancer fatality is used as the risk measure since quantitative health objectives for it have been established in the Commission's Safety Goal Policy. However, the use of 25 rem TEDE in these accident

(

dose guidelines is not intended to imply that this value constitutes an acceptable limit for emergency doses to l

the public under accident conditions. Rather, this 25 rem TEDE value has been stated in these guides as a l

e PRELIMINARY DRAFT PRELIMINARY DRAFT PRELIMINARY DRAFT reference value, which can be used in the evaluation of proposed design basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence, and hw risk of public exposure to l

radiation.

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9. Part 50, Appendix A, General Design Criterion 19, is amended to read as follows:

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Appendix A to Part 50 -- General Design Criteria for Nuclear Power Plants Criterion 19 - Controt room. A control room shall be provided from which actions can i

be taken to operate the nuclear power unit safely under normal conditions and to maintain l

it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in l

excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of l

the accident.

l Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary I

instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

i Applicants for construction permits under this part, or a design certification or i

combined license under Part 52 of this chapter who apply on or after January 10,1997, l

shall meet the requirements of this criterion, except that radiation exposures shall not exceed 5 rem total effective dose equivalent (TEDE) as defined in @50.2 of this chapter for the duration of the accident.

l PART 54 -- REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR I

NUCLEAR POWER PLANTS l

10. The authority citation for Part 54 continues to read as follows:

Authority: Secs. 102,103,104,161,181,182,183,186,189,68 Stat. 936,937,938, 948,953,954,955, as amended, sec. 234,83 Stat.1244, as amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs 201, 202, 206, 88 Stat.

1242,1244, as amended (42 U.S.C. 5841,5842), E.O.12829,3 CFR,1993 Comp., p.

570; E.O.12958, as amended,3 CFR,1995 Comp., p. 333; E.O.12968,3 CFR,1995 Comp., p. 391.

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11. Section 54.4 is amended by revising paragraph (a)(1)(iii) to read as follows:

954.4. Scope.

(a)

(1)

(iii) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in Q50.34(a)(1),

650.67(b)(2), or @100.11 of this chapter, as applicable.

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c REVIEW OF PILOT PLANT SUBMITTALS USING REVISED ACCIDENT SOURCE TERM PLANTS APPLICATION STATUS

[

(1) Perry Deletion of MSLCS Active Review Increase Allowable MSIV Leak Rate (2) Browns Ferry Deletion of MSLCS RAI Issued (Hold) 1,2, and 3 Increase Allowable MSIV Leak Rate (3) Indian Point 2 Deletion of Spray Additive Tank Review Initiated and Emergency filtration system (4) Oyster Creek Control Room Habitability Evaluation Review initiated (5) Grand Gulf (No Submittal) b i

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e C

REVIEW OF PERRY PILOT PLANT SUBMITTAL USING REVISED ACCIDENT SOURCE TERM e

Met with Perry on September 15,1998 e

Follow-up Discussion scheduled on October 8,1998 Steaming rate for first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into a LOCA Terminate fission-product release at 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (reflooding)

Homogeneous distribution of fission-product throughout drywell and containment at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Resubmittal of Radiological Consequence Calculation (October 30,1998) e Completion of Safety Evaluation (December 1998) e

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-E79 I:a l

GGNS PILOT PLANT L

ACTIVITIES 1

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i Presented at:

NEI/NRC Meeting By:

October 1,1998 Mike Withrow Rockville, MD GGNS Safety Analysis l

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I

.c GGNS RST LICENSING

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ED i

t a BWROG Gap Timing Report t

- submitted May 6,1997 on GGNS Docket i

- no formal NRC questions

- GGNS still requests NRC review and approval

- timing-only approach considered most cost-effective RST application

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i

l' GGNS RST LICENSING

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EE D

m Current GGNS Licensing Plan

- IV ake su amittal in fourth quarter 1998 qualitative, timing-only application

- Full scope submittal scope to be determined based on released NRC methods and insights from SECY-98-154

l GGNS RST LICENSING p

m Ini:ial Submittal Under Preparation

- schec uled for October 1998

- request for two minute containment isolation requirement basec on BWROG report

- calcu ation developed considering only reactor coolant source terms and no containment

<1 rem thyroid NRC rebaselining ignored reactor coolant source terms (except for MSLB)

c, GGNS RST LICENSING mus:a m Full scope submitta.

- schedulec for J~anuary 1999

- using NRC methodologies RADTRAD HABIT PAVAN 97 ARCON96

- address 3001 pH control

POTENTIAL APPLICATIONS i

M IO m Potential GGNS Full-Scope Applications

- charcoal deletion / downgrading

- manual actuation of secondary containment i

- control room boundary relaxations

- increased MSIV/ containment allowable leakrates

- CR fresh air system deletion / downgrading

- relaxed EQ dose requirements (airborne) j i

- relaxed containment requirements during fuel j

movement i

y o

Nuclear Energy institute Project No. 689 cc:

Mr. Ralph Beedle Ms. Lynnette Hendricks, Directnr Senior Vice President Plant Support

{

and Chief Nuclear Officer Nuclear Energy Institute Nuclear Energy Institute Suite 400 Suite 400 1776 i Street, NW 1776 i Street, NW Washington, DC 20006-3708 Washington, DC 20006-3708 Mr. Alex Marion, Director Mr. Charles B. Brinkman, Director Programs Washington Operations Nuclear Energy Institute ABB-Combustion Engineering, Inc.

Suite 400 12300 Twinbrook Parkway, Suite 330 1776 i Street, NW Rockville, Maryland 20852 Washington, DC 20006-3708 Mr. David Modeen, Director Engineering Nuclear Energy Institute Suite 400 1776 l Street, NW Washington, DC 20006-3708 Mr. Anthony Pietrangelo, Director Licensing Nuclear Energy Institute Suite 400 1776 i Street, NW Washington, DC 20006-3708 Mr. Nicholas J. Liparulo, Manager Nsclear Safety and Regulatory Activities Nuclear and Advanced Technology Division Westinghouse Electric Corporation P.O. Box 355 Pittsburgh, Pennsylvania 15230 Mr. Jim Davis, Director Operations Nuclear Energy Institute Suite 400 1776 l Street, NW Washington, DC 20006-3708

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. Distribution: Mtg. Summary w/ NEl re Revised Source Term Dated November 20, 1998

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Hard Cooy.

Pusuc, PERB R/F l

.OGC ACRS SMagruder SLaVie SCollins/FMiraglia BSheron BBoger JRoe DMatthews TEssig CMiller BZaleman -

REmch SLaVie -

MBlumberg JLee SMagruder DPickett CAder, RES JSchaperow, RES 1

CTinkler, RES CGingrich JMitchell, EDO GTracy, EDO l

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