ML20154C477
| ML20154C477 | |
| Person / Time | |
|---|---|
| Site: | Perry, Grand Gulf, River Bend, Clinton  |
| Issue date: | 10/02/1998 |
| From: | Donohew J NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| RTR-REGGD-01.021, RTR-REGGD-1.021 NUDOCS 9810060282 | |
| Download: ML20154C477 (28) | |
Text
- . . . . - . - - - - -- - - -- - -- - = - - - --
[ UNITED STATES i
i
- 5 2
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. maa -9 ,
?
\*****/ October 2, 1998
\'
I i LICENSEE: CLEVELAND ELECTRIC ILLUMINATING COMPANY ENTERGY OPERATIONS, INC.
i{ ILLINOIS POWER COMPANY 5
FACILITY: CLINTON POWER STATION
} GRAND GULF NUCLEAR STATION, UNIT 1
- PERRY NUCLEAR POWER PLANT, UNITS 1 AND 2
- RIVER BEND STATION 1
l
SUBJECT:
MEETING
SUMMARY
OF SEPTEMBER 9,1998, MEETING TO DISCUSS THE PLANNED JOINT PROPOSALS ON CONTAINMENT 3' REQUIREMENTS TO MITIGATE FUEL HANDLING ACCIDENTS j DURING REFUELING f
l lNTRODUCTION i
, A meeting was conducted on Wednesday, September 9,1998, between the Nuclear Regulatory Commission (NRC) staff and the licensees for the Grand Gulf Nuclear Station, Unit 1 (GGNS),
j Perry Nuclear Power Piant, Units 1 and 2 (PNPP), and River Bend Station (RBS). All of these i I plants are General Electric BWR/6 plants and are in a joint effort to propose Technical l
! Specifications (TSs) to reduce the requirements on secondary containment integrity during refueling. Although Illinois Power Company, the licensee for Clinton Power Station, did not i
. attend the meeting, it is particpating in the licensees' joint effort. The meeting was held at the i request of the licensees to brief NRC on the licensees' plans for submitting proposed TSs to
, reduce containment requirements to mitigate the fuel handling accident during refueling ,
j outages. The notice for the meeting was issued on August 28,1998. !
l
. Attachment 1 is the list of attendees, and Attachment 2 is the handout provided by the licensee
, at the meeting. Each page of Attachment 2 consists of two slides from the presentation, except i for page 12 which is a copy of the firs; page from Regulatory Guide 1.21 on measuring, i evaluating, and reporting radioactivity in solid wastes and releases of radioactive materials in
- liquid and gaseous effluents from light-water-cooled nuclear power plants. The staff did not i ;
i provide any handout at the meeting. (
i BACKGROUND j The joint effort to reduce TSs for secondary containment integrity during refueling resulted in t submittels being made by the licensees in 1994 through 1996 that proposed changes to the :
TSs for the plants. At the end of 1996, the NRC staff placed the review of the plant submittals on-hold because the work being done to finalize the draft shutdown rule was expected to address the same issues in the review of the licensees'submittals. The lead licensee for the proposed changes to the plant TSs was Entergy Operations, Inc. for GGNS. Letters dated May 24,1996, and July 16,1997, were sent to Entergy Operations, Inc. for GGNS explaining 9810060282 981002
- = ^ = " " = ' Wit FR.E CENTE COM
-r e a-M
- o l
that the staff review would remain on-hold until the shutdown rule was finalized. Because the
~ rule has been canceled, the licensees have stated that they want the staff to complete the review of their proposals.
MEETING l The agenda and purpose of the meeting are provided in pages 1 and 2 of Attachment 2.
The licensees' briefly discussed the background of the previous submittals to the NRC and the i review not completed as of December 1995. The licensees listed two amendments issued for GGNS and RBS in 1996 on technical specifications that were a part of the licensees overall proposals to the staff.
The licensees' proposals were based on the design basis accident (DBA) for the fuel handling l accident (FHA) over the core inside containment. The FFA was discussed and the slides on the FHA are in pages 5 and 6 of Attachment 2. The methodology and overview for developing 4 l
the proposed changes to the plant TSs are in pages 7 and 8. The safety aspects of the !
proposed changes are in pages 11 through 13. I The applicable regulatory requirements for the DBA and the criteria for what should be in the i plant TSs for the " primary success path" for the FHA were presented by the licensees. These I are shown in pages 14 through 18 of Attachment 2. The licensees stated that Criterion 3 of the
, Commission's final policy statement for the improved TSs was that the TSs should only include those structures, systems, and components that are part of the primary success path of the safety analysis and should not include backup and diverse equipment. The licensees stated that their current plant TSs, which are all the improved TSs, go beyond this criterion for the FHA inside containment.
The licensees stated that their proposed changes to the plant TSs are in a proposed change to the Improved Standard Technical Specifications for BWR/6s (i.e., NUREG-1434, " Standard Technical Specifications General Electric Plants, BWR/6s," Revision 1, dated April 1995) through the NRC/ Nuclear Energy Institute's Technical Specifications Task Force (TSTF) and all l of their plants have been converted to the improved Technical Specifications. The changes l discussed in this meeting are TSTF-51. The licensees also stated that the Perry plant should be the lead plant because it has a refueling outage scheduled to begin in April 1999.
l I
l I
i~
l l
,g ,
w The staff stated that the licensees submittals must clearly articulate the defense-in-depth remaining for the reduced technical specifications that they would propose and not simply rely on the Commission's final policy statement.
The licensee completed its presentation and the meeting was closea.
k . no Senior Project Manager Project Directorate IV-1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation Docket Nos. 50-416,50-458,50-461, and 50-440 '
Attachments: As stated cc w/atts: See next page ,
DISTRIBUTION:
i Hard Cooy Docket File ~ PUBLIC PD4-1 r/f J.Donohew OGC (15B18) ACRS T. Gwynn (RIV) i E-MAIL ,
EAdensam (EGA1) JHannon (JNH) CHawes (CMH2) CBerlinger (CHB) 1 DWigginton (DLW) DPickett (DVP1) REmch (RLE) RTjader (TRT)
RLobel(RML) RFretz (RXF) WBeckner (WDB)_ GHubbard (GTH) i MShuaibi(MAS 4) DJackson (DTJ)
Document Name: GG090998.MTS
_OFC PM/PD4-2 s LA/PD4-1 PM/W41 f PM/PD3-3 D/PD4-h NAME JDonoh CHawesf///// D nton DP cielt JHannon DATE T/ %5 /98 4 /M /98 7 /2 //98 $ /D /98 /0/ 1 /98 COPY- YES/NO YES/NO h/NO hSMO YES/NO OFFICIAL RECORD COPY i
6-
i i
3-The staff stated that the licensees submittals must clearly articulate the defense-in-depth remaining for the reduced technical specifications that they would propose and not simply rely 4 on the Commission's final policy statement. l i
The licensee completed its presentation and the meeting was closed.
1 (xwkdsvr Jack N. Donohew, Senior Project Manager Project Directorate IV-1 Division of Reactor Projects lil/IV Office of Nuclear Reactor Regulation
' Docket Nos. 50-416,50-458,50-461, and 50-440 ,
1 Attachments: As stated ,
1 cc w/atts: See next page
j l
.- 1 Clinton Power Station, Unit 1 lilinois Power Company cc:
Walter G. MacFarland IV lllinois Department of Nuclear Safety Senior Vice President Office of Nuclear Facility Safety Clinton Power Station ATTN: Mr. Frank Nizidlek P.O. Box 678 1035 Outer Park Drive Clinton,IL 61727 Springfield,IL 62704 Wayne Romberg Joseph V. Sipek ,
Manager Nuclear Station Director- Licensing Engineering Department Clinton Power Station j Clinton Power Station P.O. Box 678 P.O. Box 678 Mail Code V920 1 Clinton,IL 61727 Clinton,IL 61727 J
Resident inspector Leah Manning Stetzner U.S. Nuclear Regulatory Commission VP, General Counsel & Corp._ Secretary .
) RR#3, Box 229 A 500 South 27th Street I Clinton,IL 61727 Decatur,IL 62525 i
, R. T. Hill Licensing Services Manager 4
General Electric Company c 175 Curtner Avenue, M/C 481 San Jose, CA 95125 Regional Administrator, Region ill U.S. Nuclear Regulatory Commission 801 Warrenville Road
. Lisle, IL . 60532-4351 Chairman of DeWitt County clo County Clerk's Office DeWitt County Courthouse Clinton,IL 61727
- J. W. Blattner Project Manager Sargent & Lundy Engineers 56 East Monroe Street Chicago,IL 60603
Entergy Operations, Inc. Grand Gulf Nuclear Station cc:
Executive Vice President General Manager, GGNS
& Chief Operating Officer Entergy Operations, Inc.
Entergy Operations, Inc. P. O. Box 756 P. O. Box 31995 Port Gibson, MS 39150 Jackson, MS 39286-1995 Attomey General Wise, Carter, Child & Caraway Department of Justice P. O. Box 651 State of Loulslana Jackson, MS 39205 P. O. Box 94005 Baton Rouge, LA 70804-9005 Winston & Strawn 1400 L Street, N.W. - 12th Floor State Health Officer Washington, DC 20005-3502 State Board of Health P. O. Box 1700 Director Jackson, MS 39205
' Division of Solid Waste Management Mississippi Departme.nt of Natural Office of the Govemor Resources State of Mississippi P. O. Box 10385 Jackson, MS 39201' Jackson, MS 39209 Attomey General President, . Asst. Attomey General Claiborne County Board of Supervisors State of Mississippi P. O. Box 339 P. O. Box 22947 Port Gibson, MS 39150 Jackson, MS 39225 Regional Administrator, Region IV Vice President, Operations Support U.S. Nuclear Regulatory Commission Entergy Operations, Inc. l 611 Ryan Plaza Drive, Suite 1000 P.O. Box 31995 l Arlington, TX 76011 Jackson, MS 39286-1995 l Senior Resident Inspector Director, Nuclear Safety U. S. Nuclear Regulatory Commission and Regulatory Affairs P. O. Box 399 Entergy Operations, Inc.
Port Gibson, MS 39150 P.O. Box 756 Port Gibson, MS 39150 Mr. William A. Eaton Vice President, Operations GGNS i Entergy Operations, Inc.
P. O. Box 756 l Port Gibson, MS 39150 i
Centerior Service Company Perry Nuclear Power Plant, Units 1 and 2 cc:
Mary E. O'Reilly James R. Williams FirstEnergy- A290 Chief of Staff 10 Center Road _
Ohio Emergency Management Agency Perry, OH 44081 2855 West Dublin Granville Road Columbus, OH 43235-2206 Resident inspector's Office U.S. Nuclear Regulatory Commission Donna Owens, Director P.O. Box 331 Ohio Department of Commerce Perry, OH 44081-0331 Division of Industrial Compliance Bureau of Operations & Maintenance Regional Administrator, Region lil 6606 Tussing Road U.S. Nuclear Regulatory Commission P.O. Box 4009 801 Warrenville Road Reynoldsburg, OH 43068-9009 Lisle, IL 60532-4531 Mayor, Village of North Perry Sue Hiatt North Perry Village Hall OCRE Interim Representative 4778 Lockwood Road 8275 Munson North Perry Village, OH 44081 Mentor, OH 44060 Radiological Health Program Henry L. Hegrat Ohio Department of Health Regulatory Affairs Manager P.O. Box 118 Cleveland Electric illuminating Co. Columbus, OH 43266-0118 Perry Nuclear Power Plant P.O. Box 97, A210 Ohio Environmental Protection Perry, OH 44081 Agency DERR-Compliance Unit Lew W. Myers ATTN: Mr. Zack A. Clayton Vice President - Nuclear, Perry P.O. Box 1049 Centerior Service Company Columbus, OH 43266-0149 P.O. Box 97 A200 Perry, OH 44081 Chairman Perry Township Board of Trustees Mayor, Village of Perry 3750 Center Road, Box 65 4203 Harper Street Perry, OH 44081 Perry, OH_ 44081 State of Ohio FirstEnergy Corporation Public Utilities Commission Michael Beiting East Broad Street Associate General Counsel Columbus, OH 43266-0573 76 S. Main Akron, OH 44308 William R. Kanda, Jr., Plant Manager Cleveland Electric illuminating Co.
Perry Nuclear Power Plant P.O. Box 97, SB306 Perry, OH 44081 I
Entergy Operations, Inc. River Bend Station
)
cc:
I Winston & Strawn Executive Vice President and 1400 L Street, N.W. Chief Operating Officer l Washington, DC 20005-3502 Entergy Operations, Inc.
P. O. Box 31995 Manager - Licensing Jackson, MS 39286 Entergy Operations, Inc.
River Bend Station General Manager- Plant Operations P. O. Box 220 Entergy Operations, Inc. l St. Francisville, LA 70775 River Bend Station P. O. Box 220 Senior Resident inspector St. Francisville, LA 70775 P. O. Box 1050 St. Francisville, LA 70775 Director - Nuclear Safety Entergy Operations, Inc.
President of West Feliciana River Bend Station Police Jury P. O. Box 220 P. O. Box 1921 St. Francisville, LA 70775 St. Francisville, LA 70775 Vice President - Operations Support Regional Administrator, Region IV Entergy Operations, Inc.
U.S. Nuclear Regulatory Commission P. O. Box 31995 611 Ryan Plaza Drive, Suite 1000 - Jackson, MS 39286-1995 Arlington,TX 76011 Attorney General l Ms. H. Anne Plettinger State of Louisiana !
3456 Villa Rose Drive P. O. Box 94095 l Baton Rouge, LA 70806 Baton Rouge, LA 70804-9095 I Administrator Wise, Carter, Child & Caraway )
Louisiana Radiation Protection Division P. O. Box 651 l P. O. Box 82135 Jackson, MS 39205 Baton Rouge, LA 70884-2135 -
Mr. Randall K. Edington Vice President - Operations Entergy Operations, Inc.
River Bend Station P.O. Box 220 St. Francisville, LA 70775 4
e
LIST OF ATTENDEES AT MEETING OF JUNE 25.1998 GRAND GULF BULLETIN 96-03 ECCS SUCTION STRAINER
. NAME AFFILIATION
'J. Donohew - NRC/NRR/PDIV-1.
E. Adensam NRC/NRR/DRPW D. Wigginton NRC/NRR/PDIV-1 l D. Pickett NRC/NRR/PDlll-3 R. Fretz NRC/NRR/PDIV-1 C. Berlinger NRC/NRR/SCSB R.Lobel NRC/NRR/SCSB ;
R. Emch NRC/NRR/PERB '
- T. Tjader NRC/NRR/TSB
, G. Hubbard NRC/NRR/SPLB D. Jackson NRC/NRR/SPLB M. Shuaibi NRC/NRR/SRXB i
-K.Hughey EOl- Grand Gulf B. Ford EOl- Grand Gulf '
B. Burmeister EOl- River Bend B. Ferrell CEI- Perry Nuclear Power Plant W. Barber McGraw Hill where: CEI = Cleveland Electric illuminating Company EOl = Entergy Operations, Inc.
~NRC = Nuclear Regulatory Commission NRR = Office of Nuclear Reactor Regulation PDX-Y.. = Project Directorate X-Y PERB = Emergency Preparedness and Radiation Protection Branch SCSB = Containment Systems and Severe Accident Branch SPLB = Plant Systems Branch SRXB = Reactor Systems Branch '
TSB = Technical Specifications Branch ATTACHMENT 1 i
I,I l lh
[ lll l
I j! Containment Requirements l 7 to Mitigate Fuel Handling jir Accidents l'
i
- l September 9,1998 o
- i
+ ij
!l ljl Agenda ,
nj
+ Background ij
+ Analyses
+ Requested Change
+ Safety
+ Regulatory Requirements
+ Summary ATTACHMENT 2 I
1 . .
1
~
l
[
l l l ..-
- m i ;
- mo 4 r i l 9,
- il 1: !,i
!,)
1)j jl Background w .
l r.
l l 1 r-I ll>
{
l i;
j,! ,n C !! Meeting Purpose
! .o
- j + Resolve the containment requirements during fuel movement
- !!! Issue as a group of BWR 6s ai
+ Discuss the technical, safety, and regulatory basis for the roquested change l e Time is right to resolve the issue
- Extended period of time the issue has been open
- resolution of shutdown rule
- Approval of the change for permanently shutdown plants
! - Draft NUREG 1625 l
l l
ml m
hl j introduction 7 :!
h'p + Requested change has been open for a significant period of f time ( >4.5 years) without technical issues being identified
&l[
!: + Main Staff concern vo!ced (affect of shutdown rule) has been resolv.d without aff.cting r.qu.st.d change l
I!L i + Expected to save over $500K over the life of the plant at GGNS
+ Expected to save RBS and PNPP approximately 600K each per outage including critical path time
+ Change increases safety l
+ Piecemeal approach is not resource effective y!O WH mp Hy FHA Request Timeline WL
% ll + GGNS submitted originalTS change 11/94 l + BWR 6s met with staff (NRC request) 7/95
+ GGNS revised request to reflect meeting 8/95
+ RBS submitted TS change 8/95
- PNPP submitted TS change 11/95
+ BWR 6s (and others) met with staff 1/96
+ RBS and PNPP receive approval to have 2/96 primary containment airlock open 3
l .
Wl Wh
!l! ! FHA Request Timeline p, n
- h ' + Change to TS NUREGs submitted byIndustry 3/96 l [gill (NRC request)
! l!i ' + GGNS requests review prior to RFO8 5/96
'l f + NRC identifies they are unable to compete review 6/96 Ii but that the review should be complete by 9/97 r
+ GGNS meets with the Staff to discuss onetime 7/96 J TS change to allow repairs
+ Staff grants onetime GGNS TS change 10/96
+ Draft NUREG 1625 proposed similar requirements 3/98 i and identified approval for Trojan I
4 q..
t r
{hlJ Analyses pu 4
I l
k
1
. 1 YI I i Current License Basis Lj FHA Analysis !
@U y 1 + Two analyzed events jh - Fuel Handtlag Accidentin the Auxiliary Building gn UFSAR Se etion 15.7.4 l l - Fuel hanc ling accident in the primary containment UFSAR S. !ction 15.7.6
- The secondary containment (i.e., auxiliary building and enclosure building) working in conjunction with the SGT System limit the radiological consequences (per SRP 15.7.4 guidelines) to well within the 10CFR100.11 limits (i.e.,75 rem thyroid and 6 rem whole body).
I il L' 'I n:
llll l Reanalyzed FHA l
+ Does not credit the active engineered safety feature (ESF) l" systems (e.g., auxiliary building and enclosure building
! Integrity, Isolation of the containment and fuel handling area ventilation systems, and the SGT System) that are currently credited in the UFSAR analyses to reduce the consequences of the analyzed events
+ Otherwise assumptions are consistent with the analysis
_ presented in UFSAR Sections 15.7.4 and 15.7.6
+ Demonstrated that the dose limitations of SRP 15.7.4 are satisfied for decay periods of 12 days or more without credit for the ESF systems b
N ui 19 n!i Analysis Summary 1:' :
I + Following radioact!<e decay, ESF Systems are not required gi during a fuel handling accident to maintain calculated doses hJ less than the regulatory guidance (e.g.,75 rem thyroid offsite lll and 30 rem thyroid control room).
j + The results of these analysis were submitted November 9, i
E"j 1994 and discussed in subsequent meetings.
+ Staff has performed independent calculations.
+ Calculations formed the basis for the onetime TS change r approval.
- The acceptability of the calculations is not an issue between the licensee and the Staff.
i.
f
/l i pI Nll t: ;
ai H
Requested Change et 6
'I j .. Methodology for Developing the
[h Requested Changes qi !l + Follows the guidance of the Rulemaking on TS Improvement,
<hh by focusing the TS requirements on those systems necessary llJ to mitigate postulated events I!i{
! f h + Recognizes that radioactive decay is an effective means of f mitigating an FHA l !.
!
- Recognizes that the only CORE ALTERATION postulated to O result in fuel damage is an FHA
,
- Retains the requirement for OPERABILITY of systems used to mitigate the dose consequences of an FHA during the time
,i frame the analysis takes credit for their functioning s
n
- I f
t J, Overview of Proposed Technical Ln Specification 7i,;
- Retains the requirement for OPERABILITY of systems used to mitigate the dose consequences of an FMA during the time l'j lL Ji frame the analysis takes credit for their functioning
_ + Does not alter the TS requirements concerning operations with potential for draining the reactor vessel
+ Does not alter the TS requirements for protection from
_ criticality events
+ Does not alter the TS requirements for decay heat removal l
l l
[
7 . _
p '.
P.
+ .
Details of l lc- Proposed Change l
- ..; i l 1h , + Requires dose mitigation systems to be OPERABLE when ill .j handling "recently irradiated fuel assemblies'"
I) .
i
- + Removes the requirement for dose mitigation systems to be j 'I J OPERABLE during CORE ALTERATIONS
,i
[
- Provides Bases discussions describing the relevant limit l
l l
l l
l l
i!' .
m h
h! l lit d
l lb Safety Aspects of the
_ Proposed Change W
l H :i gn Overall Outage Safety F l; H U + Outage Tech Specs are based on specific events (e.g. FHA, l
!! i draindown) not overall shutdown safety considerations b,l ll
> + As a result, current Tech Specs force some ou ll Into relatively higher risk periods 1
+ Proposed changes result in overall outage safety gain
- flexibility to schedule activities when mitigate resources most available
- fuel movement conducted during relatively lower risk periods Relationship of Fuel Handling to Water Level and Shutdown Risk 1a , .
140 taeveman M-te i 120 p gg,9g g 100 h 80 1E 11 8
m 40 L***8 ",1E 12
{ , '
Leur Water " ' '
y ym W 1E-13 0 5 10 15 N 25 30 Days Following Shutdown
1 oe H
0, Decay Time and Water Level mn J d + Original analyses assumed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> decay time i: h (Il + Enforcing a decay period is a more reliable " defense in depth" mechanism than traditional physical barriers l
[;;
~
j r + Decay is an appropriate means of mitigating the effects of a l lij fuel handling accident, substituting for less reliable features
- Similarly, water level is a " defense in depth" barrier not
~
l usually applied to other accident classes, and is specifically controlled in Tech Specs for the fuel handling accident
+ Decay and water level are performing the " defense in depth" containment function of dose mitigation
, I l
. In h;h hjh Dose H I' j
- Licensing basis dose calculations meet regulatory safety
.j; guidance (i.e., SRP acceptance criteria) and are well below
.!L regulatory requirements l
__ + Independent NRC calculations concur
+ Realistic fuel handling accident estimates are 13 orders of magnitude lower
. 10
l p ! Risk Perspective I II + From Grand Gulf shutdown risk studies
[ !I Fuel Handling Core Damage O]I:
Accident Event
'l '
ji I l l Frequency (/RY) 7 X 104 1 X 104 Release (Cl) 81 3.35 X 108
- (1131 equivalent)
Risk (Cl/RY) .006 3.35 Safety focus should be on shifting activities to low probability CDF periods H
i.
fi !!
I; .i
+n Containment Closure il O
+ Closure will mitigate a realistic fuel handling accident release mi + BWR6s each have commitments to plans and administrative l- controls to ensure containment closure l + Design differences lead to differences in the meaning of
" closure"
~
- Grand Gulf /Clinton secondary containment (primary containment hatch closure not required during shutdown)
- River Bend / Perry - primary containment
+ GDCs, Part 20 and ODCM require releases to be monitored and controlled
+ Proposed Industry maintenance rule guidance requires closure capability to be available II
Reviion 1 U.S. ATOMIC ENERGY COMMISSION I
(,,
REGULATORY GU DE DIRECTORATE OF REGUL.ATORY STANDARDS REGULATORY GUIDE 1.21 MEASURING, EVALUATING, AND REPORTING RADIOACTIVITY IN SOLID WASTES AND RELEASES OF RADIOACTIVE MATERIALS IN LIQUID AND GASEOUS EFFLUENTS FROM LIGHT. WATER-COOLED NUCLEAR POWER PLANTS A.INTROD**OTsoN days after January I and July I of each year which spedfies the quantity of each of the principal General Design Criterion 60, " Control of releases radionuclides released to unrestricted areasin liquid and of radioactive materials to the environment," of in gaseous effluents during the previous 6 months of Appendix A, " General Design Criteria for Nuclear Operation, and such other information as may be Power Plants," to 10 CFR Part 50, "!) censing of required by the Corrmission to estimate maximum Production and Utilization Facilities," requires that the Potential annual radiation doses to the public resulting nuclear power plant design include means to control from effluent releases, the release of radioactive materials in gaseous and liquid effluents and to handle indioactive solid wastes Paragraph (c) of g20.I, " Purpose," of 10 CFR Part produced during normal reacter operation, including 20 states that every reasonable effort should be made by anticipated operational occurrences. AEC licensees to maintain radiation exposure, and l releases of racioactive materials in effluents to General Design Criterion 64, " Monitoring unrestricted areas, as far below the lirnits specified in r radioactivity releases," requires that nuclear power Part 20 as practicable, i.e., as. low u is practicably
( plant designs provide rneans for monitoring effluent discharge paths for radioactivity that may be released achievable, taking into account the state of technology, and the economics of improwments in relation to from normal o pera tions, including anticipated benefits to the public health and safety and in relation operational occurrences, and from postulated accidents. to the utilization of atomic energy in the public interest.
Section 20.106, " Concentrations in effluents to This guide describes programs acceptable to the unrestricted areas," of 10 CFR Part 20," Standards for Regulatory staff for measuring, reporting, and evaluating Protection Against Radiation," provides that a licensee releases of radioactive materials in liquid and gaseous shall not release to an unrestri:ted ares, radioactiw effluents and guidelines for classifying and reporting the materials in concentrations which exceed limits categories and curie content of solid wastes. Other specified in 10 CFR Part 20 or as otherwise autho. zed programs for the reporting of operating information, in a license issued by the Comnussion. Section 20.201, including abnormal occurrences, are presented in
" Surveys," of 10 CFR Part 20 further requires that a Regulatory Guide 1.16. " Reporting of Operating licensee conduct surveys of concentrations of Information." la some cases, specific programs should radioactive rnaterials as necessary to demonstrate be supplemented because of individual plant design compliance with AEC regulations. features or other factors. The need for supplemental or modified programs will be determined on a case.by. case Paragraph (a)(2) of 6 5 0.36a, " Technical basis.
specifications on efTluents from nuclear power reactors,"
of 10 CFR Part 50 provides that technical specifications The Advisory Committee on Reactor Safeguards has for each license will include a requirement that the been consulted concerning this guide and has concurred licensee submit a report to the Commhsion within 60 in the regulatory position.
" " " * " " " 24- ".airr,.T,:l"a *#rd *r".' ::llll:'o*1':::t
- ,,0ll:*.l':fc :,3:".,*'#,':*.' :::'lll '."2',20 e--
.= : =. s:: i "t?,w*p"=.""lll"J.*:";-l7
, lm. - *::::: :::::-l%:" %
O '"*
a:::' 2 == =: _-~: .
- 7:r.'..'* T.::L*::::::,':,,2 :::1-ll'::::
- ::::.':" " ""'"",' ",.*.",r;*a:.*.t'Or_' a:'2' - "
n.
e-i ,"::l".l:,T L .", . '. I o..'.""".'".
" W:: a" .' ' .
.*.,*-P.,;."::""= " i' ,2::: .1 o"" .- ~
1 R.
'i l
l l i
!! Safety Conclusions !
I h1 iii l
- Proposed changes lead to a not outage safety benefit - i lji j reduction in activity during relatively higher risk periods '
l
&ll -
- Fuel handling accident safety criteria are all met I
+ Fuel handling accident " defense in depth" barriers are
~
different (but as effective) as barriers for other accident classes l
pl!
l l lll ll Regulatory
_ Requirements 14 13
a
! l Applicable l;l h Regulatory Requirements
!i ll
[ !, + 10 CFR 100
- Limits offsite doses to < 300 rem thyroid l l + 10 CFR 50.36
]'j f' - Requires Technical Specifications be established and maintained and identifies the requirements to be included
+ Standard Review Plan 15.7.4
- Limits offsite doses to well within the 10CFR100.11 limits (i.e.,75 rem thyroid and 6 rem whole body).
1 -
i l
n N, Why Criteria Were Developed l lllj for the Technical Specifications l
The Technical Specifications had become so controlling of all
.- aspects of plant operation that unneeded requirements in the Technical Specifications were diverting both staff and licensee
_ attention from the more important requirements to the extent that the excessive requirements have "resultedin an adverse but unquantifiable impact on safety".
~, ._ _ . . . _ . __ _ . _ _ _ _ _ . _ _ . _ _ _ _ . _ . _ _ _
e L 10 CFR 50.36 Criteria For d ll Technical Specifications lo ll n a Criterion 1: Instrumentation that is used to detect, and indicate i j' in the control room, a significant abnormal degradation of the p__i! reactor coolant pressure boundary f al Criterion DBA or Transient 2: A Analysis process that variable that either assumes the is anofinitial failure or condition of a
};j presents a challenge to the integrity of a fission product r; barrier l Criterion 3: An SSC that is part of the primary success path and y which functions or actuates to mitigate a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fiscion product barrier i Criterion 4: A structurc, .ystem, or component which operating i experience or PSA has scown to be significant to public
! health and safety 4 1
- l. :l
- u. -
k il 10 CFR 50.36 Rulemaking 1,
!~ "If a technical specification provision does not meet any of the first three criteria, and if the current PRA knowledge or operating experience does not identify the structure, system, or component as risk significant, the NRC staff will not
~
preclude relocating such technical specifications."
l l
l . .
o i f j Technical Specification l
! Improvement Criterion 3 li , 1 0
L ;
El A structure system, or component that is part of the primary
?~ success path and which functions or actuates to mitigate a l
! ~
Design Basis Accident or Transient that either assumes the '
failure of or presents a challenge to the integrity of a fission l product barrier.
1
- l (l! The Final Policy Statement yil Concerning Criterion 3
(. i L! !
4 !
l] The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate ( including consideration for single failure criteria), so that the plant response to the DBA and Transients limits the consequences of these events to withle .he appropriate acceptance crfteria. It is the intent c. als criteria to capture into the TS only those SSCs that are part of the primary success path of the safety analysis... The primary success path does not include backup and diverse equipment .
9 a
ui Staff Position on " Appropriate n .i Acceptance Criteria" h
i I;l lj; e The definition discussed in T. E.ofMurley's
" appropriate acceptance letter dated criteria"was clearly May 9,1988. This IQl letter transmitted the results of the staff's review of the Owners Groups' application of the TS selection criteria and
{'
- __I formed the basis for the issued Improved Standard TS, d
+ Enclosure Section 2.(6) states:
"Accordingly, the SRP limits should be used to define the equipment in the primary success path for mitigating accidents and transients when developing the new STS."
I h
c li iu pp y jj Permanently Shutdown Plants i !
ji ' + Draft NUREG 1625 identifies that TS Containment f-J requirements are not required when dose analysis no longer credits their functioning.
i
+ Permanently shutdown plants have been licensed allowing fuel movement without TS Containment requirements.
i i
t
\7
?
1 .
.c jit! Comparison to Similar Changes M .
Approved by the NRC l
! ll jj yll+ Generic PWR effort to reduce containment requirements l h during fuel handling.
JI i
+ Approved changes include plants relying on operator actions and the ability to restore containment to protect small fraction dose limit.
+ Permanently shutdown plants have been licensed allowing fuel movement without TS Containment requirements.
+ BWR 6 proposal retains all OPERABILITY requirements for equipment during the time frames that the equipment is needed to protect small fraction dose limit.
I t1 ih Regulatory Requirements i;! Summary
!L lll
- Requested changes are in accordance with established Staff positions
+ Retains the requirement for OPERABILITY of systems used to mitigate the dose consequences of an FHA during the time frame the analysis takes credit for their functioning
+ Allows plant staff to schedule activities during most cost effective and least risky time frames l
i l
l lB
, s
, II, it. I 1H l I
1 '!
P iW i Summary I ,_
ll t,
l i
'I
.i. :
[l Summary iii l
[' + Overalllevel of safety improves through implementation of the j proposed changes m
] + Defense in depth is preserved
- + Requested changes are in accordance with the Technical Specification improvement Rule
+ Retains the requirement for OPERABILTTY of systems used to
_ mitigate the dose consequences of an FHA during the time frame the analysis takes credit for their functioning
- Regulatory requirements are met