ML20212E301
| ML20212E301 | |
| Person / Time | |
|---|---|
| Issue date: | 10/30/1997 |
| From: | Coyne K NRC (Affiliation Not Assigned) |
| To: | NRC |
| References | |
| NUDOCS 9711030137 | |
| Download: ML20212E301 (70) | |
Text
-.
4k g20tg
+
Y t
UNITED STATES j
s Ij NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 20066-0001
\\...../
october 30. 1997 n
MEMORANDUM TO:
Public Document Room FROM:
Kevin Coyne Probabilistic' Safety Assessment Branch Division of Systems Safety and Analysis Office of Nuclear Reactor Regulation.
SUBJECT:
RISK-INFORMED REGULATION SEMINAR On October 28,1997, the Director of the Division of Systems Safety and Analysis, Office of Nuclear Reactor _ Regulation (NRR), lead an overview seminar on risk-informed regulation for the NRR technical staff. The seminar provided a brief background on the PRA policy statement, the scope of risk-informed regulation, staff expectations, staff responsibilities and proposed acceptance criteria. The viewgraphs used during the presentation are included in the Attachment to this memorandum.
Attachment:
. As stated Distribution (w/ attachment):
-)
File Center
)\\
. [O
. SPSB File PDR
\\
KCoyne MRubin JFlack
@NJ/
1)yph Ith gfd
as
.- ~~
,_c
'#[,
gaN*
b,fu""~
g gjgggggjjp 9711030137 971030 PDR ORQ NRRA PDR
.ww.
y me. -,aia.wm--w..a pss 64.-
4..,.
4 se%.W.4. age
..M,.D
- 5.eres.A..d'_D+.mm4._4,uu+#_Mm.m.
4Maebhd-.*u m n.ws m.v.<M_e,,4e-es'n4Ee.M weJadaah. O m 1-AFA4.hm_
5 4
'.g' 4
i i
7 R
seenZ
\\'
mO Q
STATg 3m 8,9 i
g gm 4.$p/ %s, Ob
\\
"D $
- $, S J
h i
i oO
/WnoO h c'
>f.
l i
O l
2 a
C r
k a
~'
s i
4 8
--s-.u+.
e
..4aa.
ea-~~---
na-u nn a am w-a-.-
am-u a, n um u.w m-+-
--.x.
mm-w-r---an,-.--
-s.*am----n-a
-e - m ee+-ap
+-,n-
..- - - - _-- g a
i i.
l i
I l
1 i
I I
r 9
9 9'
i o
e a
V)
O i-2 V
A" o
C 8
g a
i 7
c 0".
lll3
- n m.mm a mmu W
)
i o.e.
e
-me s mene O
i U) 4 j'
k, f
ii
)
3
)
i
)
i i
e
i PRA POLICY STATEMENT The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.
PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal of additional regulatory requirements in accordance with 10 CFR 50.109 (Backfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent o'f-this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.
t PRA POLICY STATEMENT (cont.)
PRA evaluations in support of regulatory decisions should e
be as rea!istic as practicable and appropriate supporting data should be publicly available for review.
The Commissien's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgements on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.
e 4
i I
i i
RISK-INFORMED REGULATION i
insights derived from probabilistic risk assessments are used in combination with deterministic system and er gineering analyses to focus licensee and regulatory i
attention on issues commensurate with their importance to safety.
l r
l 5
OBJECTIVES FOR RISK-INFORMED REGULATION
- ]
Enhance safety decisions (e.g., configuration control, accident management)
Efficient use of NRC resources (e.g., IPE insights, risk-based inspections)
Reduced industry burden (e.g., graded QA, risk-based ISTi e
l~
s
PRA STRENGTHS i
?
Integrated and Systematic Examination of Design and Operations Features i
incorporates System Interaction and Human-system Interface e
Provides Model for incorporating Experience with the Engineered System Process for Evaluating Unceitainties in the Decision Making Process j
i e
Permits Analysis of Competing Risks Permits Analysis of New issues Via Sensitivity Studies l
l Provides insights on Relative importance of Systems, l
l l
Components, etc.
Provides Quantitative Measure of Overall Risk Associated l
with Engineered Systems
[
l I'
7 l
~
PRA LIMITATIONS Excludes Sabotage Acts of Commission Transition Risk Aging and Degradation Equipment Survivability System Interaction Management Culture Manufacturing Defects l
I!!iIIiII e
uI aE n
pf nn do dc O
ar Ro V
t m
em e
gp E P s s aC o
R R ia s
no n s AA dm s e bm s
Li LM r
a i i es c
P f s
.t O
i L
i i
n o v
gn BE i
t s o e
J i
M f
s E
s t
n C E i
f N
T N a
8 T
f I
R p
,R ET r
o g
R SIO re E
s
,S AN s
A N
v P
E D
i O
L a
u
,D SA q
a CN N
e M
O-r t
r S
P ly
,S E
- I t.
h
!ll l!
!I hl
PRAIMPLEMENTATION PLAN Develop SRPs for risk-informed regulation Pilot applications for risk-informed regulatory initiatives Inspections Operatorlicensing Event assessment Evaluate use of PRA in resolution of generic issues Regulatory effectiveness evaluation Advanced reactor reviews t
Accident management Evaluating IPE insights to determine necessary follow-up activities
~
s l
RISK-INFORMED REGULATORY GUIDES
~
i
~
AND SRPS DG-1061 - General SRP Chapter 19 - General Guidance to licensees guidance to staff DG-1062-Application-SRP Section 3.9.7 -
specific guidance on application-specific in-service testing guidance on IST DG-1063 - Application-SRP Section 3.9.8 - application-specific guidance on specific guidance on ISI-in-service inspection (ISI)-
under development under development DG-1064 - Application-Inspection guidance-specific guidance on graded under development quality assurance 1
l
. DG-1065 - Application-SRP Section 16.1 - application-l specific guidance on specific guidance onTS technical syx:ifications j
10
d J
PRINCIPLES OF RISK-INFORMED REGULATION THAT GOVERN CHANGESTOTHE CURRENT LICENSING BASIS (CLB) r Maintain Defense-in-Depth r
Meet Current Maintain Sufficient Regulations Safety Margins
\\
) (
integrated Decisiomnaking implementation Proposed increases in risk and Monitoring and their cumulative effect Strategies Which are small and do not cause Address Uncertainties the NRC's Safety Goals to be exceeded a
~
t
.11
REGULATORY GUIDES AND SRPS (cont.)
Principles of Risk-informed Regulation for CLB changes The proposed change meets the current regulations.
This principle applies unless the proposed change is explicitly related to a requested exemption or rule change (i.e., a 50.12
" specific exemption" or a 2.802 " petition for rulemaking").
Defense-in-depth is maintained.
Sufficient safety margins are maintained.
Proposed increases in risk, and their cumulative effect, are small and do not cause the NRC Safety Goals to be exceeded.
Performance-based implementation and monitoring strategies are proposed that address uncertainties in analysis models and data and provide for timely feedback and corrective action.
~
12
REGULATORY GUIDES AND SRPS (cont.)
Expectations The overall risk management approach 'uses risk analysis to improve operational and engineering decisions brcadly and not just eliminates requirements the licensee sees as undesirable.
The acceptability of proposed changes should be evaluated by the licensee in an integrated fashion that ensures that all principles are met.
Core damage frequency (CDFl and large early release frequency (LERF) can be used as suitable metrics for making risk-informed regulatory decisions.
I Increases in estimated CDF and LERF resulting from proposed CLB changes will be limited to small increments.
The scope and quality of the engineering analyses (including traditional and probabilistic analyses) conducted to justify the proposed CLB change should be appropriate for the nature and l.
scope of the change and should be based on the as-built and as-l operated and maintained plant.
~
'3
i 1
REGULATORY GUIDES AND SRPS (cont.)
1 Expectations (cont.)
)
Appropriate consideration of uncertainty is given in e
analyses and interpretation of findings.
)
l j
The plant-specific PRA supporting licensee proposals has e
I been subjected to quality controls such as an independent peer review.
Data, methods, and assessment criteria used to support e
regulatory decisionmaking must be scrutable and available for pubhc review.
l' 14
PRINCIPAL ELEMENTS OF RISK-INFORMED PLANT-SPECIFIC DECISION MAKING M
PRA Traditional Analysis f
/>
x 2
N l
ij
/
N I
'I
/
s/ j'/~
II"'
N Perform Submit Implementation /
Proposed Define Engineering Moniton,ng Change Analysis Change l
Program Element 1 Element 2 Element 3 Elennnt 4 i
15
I REGULATORY GUIDES AND SRPS (cont.)
Acceptance Guidelines e
Defense-in-depth is maintained l
a reasonable balance prevention of core damage, l
prevention of containment failure, and consequence mitigation is preserved over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided system redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences of challenges ~to the system (e.g., no risk outliers) defenses against potential common cause failures are preserved and the potential for introduction of new common cause failure mechanisms is assessed independence of barriers is not degraded defenses against human errors are preserved 16
i REGULATORY GUIDES AND SRPS (cont.)
Acceptance Guidelines (cont.)
Sufficient safety margins are maintained codes and standards or alternatives approved for use by the NRC are met safety analysis acceptance criteria in the current licensing basis (e.g., FSAR, supporting analyses) are met, or proposed revisions provide sufficient margin to account for analysis and data uncertainty e
~
17
Mean Core Damage Frequency Acceptance Guidelines NOT ALLOWED LL Q
m = ~ m e m =,2 m.== = w., = = x =,1 m ~,=: p,,we c
10
- s-"
- ++-'
?
f f
a,ib'&hlI?[:jf"" ' ";' -?lW?%Y?l.'?{$f'jf,f'n')ff/$(m?d?!/?l,flk'${'jk'(pj$ mas e
nn'flYbinj+f,
- Nf?SY na apw y,
481ncreas6d Man;agementlAttentionIO e
p aw a
w%p?y$,:(al~1'4%nw,,WN?s%2~fw,kxsae, kgw'p,;g%,'nfy Wg&: m(f 1;pGyhE' nznes,C6., ish~ ww, g 4l
. -,~
Q't
~
x a
a su nnu ggp
, rw m#_,,. -
m,
... =
- -... ~
,,= ;
[,,, g*,
E 10 L
,,.,;-...e.- -
~.
7,. ~ y s
. ' 555s '
u MkMi Normal Risk-informed -
.3
$g m;$W n ME)$$$[i.NIN!!!?
M$5 Staff Review 4*
2 10 10'
~
Mean Core Damage Frequency (CDF) 18
4 1
i increase in LERF
{
[
d d
E o
8 m
e i
M"v~S)l 2j CD l
g z
O e
w
- g g'*nR3 Mi p
u O
g gi osq 3
ikif'%g'f$*
g d.e i
t i
g M
g iG k @I a ;,r
[
g-4
.e j
$s M t!
{
F 4
3.
+
me s '
a; o
=5 W
- n :a L, g h 4 4,$
sr 9
g
/
Mj :
4 O
E Y
~
dp ti
.3
(
y 8
? idk3EM yli o
m
(
'it a m M i
r
' h$i f ! n.
f 5
l m
N d
hN a
e i
e w %~q m,.
g 1<
i g pe%
i xa
=. a gg :q g c
gu 3.
l me
[
Z 3
9
- .f[{ %lg '!) -}
!MMER@j{y{i!E}j5 O
l O
O M E
. m.
m l=
l 1
2 2
l 8
O j.
m o
4 m
O
.,e o
e e
m E
T Ox e9 g n
.O m
l N
C i
71 w
1 l
)
i o
i M
i i
i i
1 REGULATORY GUIDES AND SRPS (cont.)
Acceptance Guidelines (cont.)
e For a plant with a mean core damage frequency at or above 1E-4 per reacte year (the Commission's subsidiary core i
damage frequency objective) or with a mean LERF at or above 1 E-5 per reactor year, it is expected that applications will result in a net decrease in risk or be risk neutral.
For a plant with a mean core damage frequency of less than 1 E-4 per reactor year, applications will be considered which, when combined with the LERF guidelines described below:
i i
Result in a net decrease in CDF or are CDF-neutral; or l
Result in increases in calculated CDF that are very small (e.g., CDF increase of less than 1 E-6 per reactor year); or j
l-l l
. ?--
~
REGULATORY GUIDES AND SRPS (cont.)
Acceptance Guidelines (cont.)
Result in an increase in calculated CDF in the range of 1E-6 to 1E-5 per reactor year, subject to increased NRC technical and management review and considering the following factors:.
The scope, quality, and robustness of the analysis (including, but not limited to, the PRA), including consideration and quantification of uncertainties; The base CDF and LERF of the plent; The cumulative impact of previous changes (the licensee's risk management approach);
Consideration of the Safety Goal screening criteria in the staff's Regulatory Analysis Guidelines, which define what changes in CDF and containment performance would be needed to consider potential backfits; The impact of the proposed change on operational complexity, burden on the operating staff, and overall safety practices, and Plant-specific performance and other factors, including, for example, siting factors, inspection findings, performance indicators, and operational events.
A plant with a mean CDF in the range of 1E-5 to 1E-4 and an increase in CDF of e
up to 1E-5 will be subject to increased NRC technical and management review, as described above.
'~
AND
~
21
REGULATORY GUIDES AND SRPS (cont.)
Acceptance Guidelines (cont.)
For a plant with a mean LERF of between 1E-6 and 1E-5 per e
reactor year:
Result in a net decrease in LERF or are LERF-neutral; or Result in an increase in calculated LERF of up to 1E-6 per reactor year, subject to increased NRC technical and management review, as described above; OR For a plant with a mean LERF of less than 1 E-6 per reactor year:
Result in a net decrease in LERF or are LERF-neutral; or Result in increases in calculated LERF that are very small 1
(e.g., LERF increase of less than 1 E-7 per reactor year); or l
l Result in an increase'in calculated LERF of up to 1E-6 per l
reactor year, subject to increased NRC technical and management review, as described above.
22 l
STAFF AND MANAGEMENT RESPONSIBILITIES Ensure that licensing submittals are identified and processed in accordance with risk-informed guidance; Identify current requirements that could be significantly enhanced with a risk-informed and/or performance-based w >ach Ensure the objectives of risk informed regulation are met:
Enhance safety decisions Efficient use of NRC resources Reduced industry burden Ensure adequate staff training on the use of risk-informed guidance and underlying PRA technical disciplines
~
l
STAFF AND MANAGEMENT RESPONSIBILITIES NRR Projects Initial review of submittals Determine whether previously reviewed (consult with primary review branch-PRB)
In general, unless a previous precedent exists, review by PRB required Forward to PRB for action; copy to SPSB for information For particularly complex issues and first-of-a-kind reviews,'DRP and ADT should meet to consider establishing r. review team for the project 24
~
c DECISION LOGIC FOR SUBMITTAL REVIEWS Changes to the Current Licensing Basis 9
--l-r
- ...,c Staff Propos'[es increased Requirements - Useli0.109 Back0t Rule
~
(Regulatory Analysis Guidelines):
u
.n y
a ny h
m m
- Licensee Makes >
" CURRENT LICENS98G BASIS" :..
> Change Consistent Twith 50.59 Process
...c
~ %^%
i
-Licensee Respsests Ch5nge Licensee Requests Change
... In Requirements vis Nin W seyond
~
~ Approved Staff Positions Approved Staff Positions-a
?(10 CFR 50.91H12)n -
- (10 CFR 50.91H12)l Y
Licensee Requests *
- mw
=
Does Not Present -
' ChangeConsistentwith';
Riskinformation t
Riskinformation "Use Risk informed
- Approved Staff Position.
hMW a ;(Rule, RG, SRP, BTP...)l -
RG/SRP" -
' "NormalSt.aff Review",i
STAFF AND MANAGEMENT RESPONSIBILITIES e
Primary Review Branch Review application and determine level of PRA support needed; considering:
Complexity Scope Risk Impact (perceived or calculated)
Relationship to previous reviews Existence of SRP/RG Need to rely on general RG for guidance Define necessary support reviews Integrate SER inputs and issue to PM 26
1
!~
i STAFF AND MANAGEMENT RESPONSIBILITIES 1
PRA and Severe Accident Branch Support Provide support to PRB for applications which make use of probabilistic risk analysis as part of the license amendment justification Provide guidance as necessary for applying the general RG to new applications j
Consult with SRXB or Analytical Support Group as needed for success criteria Prepare SER input to PRB t
i-i 27
l i
STAFF AND MANAGEMENT i
RESPONSIBILITIES 1
Introductory Staff Training RG and SRP Initial orientation (this seminar) transmit guidance to staff with plans and expectations issue Office Letter for risk-informed licensing reviews e
PRA for Regulatory Applications (P-105)
PRA for Technical Managers (P-107)
Additional Training (under development) f Inspection Activities 2 week course (P-111: PRA Tech. & Reg. Perspectives) emphasis on practicality by using case studies cover inspection activities of Region and NRR cover revised inspection guidance on use of PRA j
Other in-depth PRA technical training i
NUREG/BR-0228, Level 1,2,3 2s
PRA IMPLEMENTATION PLAN KEY NEAR-TERM ACTIVITIES lssue Regulatory Guides and Standard Review Plans IRES /NRR)
Develop New Application-Specific Guidance for Steam Generator PerformanceI NRR/RES)
Complete PRA Pilot Reviews I NRR)
Issue Risk-informed Inspection Guidance (NRR)
Train Staff on Use of Risk-informed Guidance (NRR/AEOD)
Address Remaining issues in Commission SRMs (RES/NRR/AEOD)
PILOT APPLICATIONS OF PRA GQA i: South Texas, Palo Verde, Grand Gulfi e
Risk-Informed IST (Palo Verde, Comanche Peak) e e
Risk-informed ISI1 Surry, ANO, FitzPatrick, Vermont Yankee)
Risk-informed Tech Specs (CEOG) e e
n
OBJECTIVES AND EXPECTATIONS FOR USE OF PRA INSIGHTS IN INSPECTION ACTIVITIES Objectives:
Integration of Risk Information with other non-PRA e
factors help to Flan Inspection Activities Evaluate the Significance of Inspection Observations and Findings
~
Evaluate the Adequacy of Licensee Propossis or Programs which have Probabilistic Risk Elements in Their Bases
~
31
OBJECTIVES AND EXPECTATIONS FOR USE OF PRA INSIGHTS IN INSPECTION ACTIVITIES (cont.)
Expectations:
Use of PRA Complements Traditional Engineering Analysis Consistency in the Use and Communication of Risk Related information within the inspection Program PRA results do not provide the sole basis for imposing Regulatory Requirements.
Quantitative thresholds alone do not determine acceptability PRA calculations can potentially result in a Large Resource impact to Licensees Inspection Activities should focus on the Dominant Accident Sequences and Success Criteria
- ' Utilize PRA Information to Characterize the Significance of Inspection issues.
32
ISSUE RISK-INFORMED INSPECTION GUIDANCE IMC 2515 Appendix C (complete) 50.59 Evaluations l'12/97)
Regular Maintenance Observations (12/97)
GQA Inspection Procedure (draft)I:12/97)-
e GQA Inspection Procedure (final) i 3/98) e I
w W -.
A RE:
j
't UNITED STATES
'g j
NUCLEAR REGULATORY COMMISSION k *. ** p #
WASHINGTON, D.C. 30666-0001 8
October 30, 1997 MEMORANDUM TO:
Public Document Room FROM:
Kevin Coyne
~#
Probabilisti afety Assessment Branch Division of Systems Safety and Analysis Office of Nuclear Reactor Regulation
SUBJECT:
RISK-INFORMED REGULATION SEMINAR On October 28,1997, the Director of the Division of Systems Safety and Analysis, Office of Nuclear Reactor Regulation (NRR), lead an overview ser.tinar on risk-hformed regulation for tha NRR technical staff. The seminar provided a brief background on the PRA policy statement, the scope of risk-informed regulation, staff expectations, staff responsibilities and proposed acceptance criteria. The viewgraphs used during the presentation are included in the Attachment to this memorandum.
Attachment.
As stated Distribution (w/ attachment):
File Center SPSB File
.PDR KCoyne MRubin JFlack w
,.m n
~
.-+.--v-~
n ATTACHMENT AB REG S'
4h 4
i s
i r
s li j
s a
+9 ll 90 n+++
OVERVIEW OF RISK-INFORMED REGULATION i
~~
1 I
i j
ki
- Background j
9
~
- Responsibilities
- Key Near-Term Activities 9
9 e
2
-tg gui.
PRA POLICY STATEMENT The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.
PRA and associated analyses (e.g., sensitivity studies, uncertainty e
analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal of additional regulatory requirements in accordance with 10 CFR 50.109 (Backfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed. It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.
8 l
PRA POLICY STATEMENT (cont.)
PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.
The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgements on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.
4
RISK-INFORMED REGULATION Insights derived from probabilistic risk assessments
~
are used in combination with deterministic system and engineering analyses to focus licensee and regulatory attention on issues commensurate with their importance to safety.
t e
____.___.-___-----2------
~
l
~
~
OBJECTIVES FOR RISK-INFORMED REGULATION Enhance safety decisions (e.g., configuration control,
)
accident management) l Efficient use of NRC resources (e.g.,IPE insights, risk-based inspections)
Reduced industry burden (e.g., graded QA, risk-based ISTi i
~
PRA STRENGTHS F
Integrated and Systematic Examination of Design and Operations Features incorporates System Interaction and Human-system Interface e
Provides Model for incorporating Experience with the Engineered System Process for Evaluating Uncertainties in the Decision Making Process
~
~
Permits Analysis of Competing Risks Permits Analysis of New issues Via Sensitivity Studies Provides insights on Relative importance of Systems, Components, etc.
Provides Quantitative Measure of Overall Risk Associated with Engineered Systems e
PRA LIMITATIONS e
Excludes Sabotage e
Acts of Commission e
Transition Risk e
Aging and Degradation Equipment Survivability e
e System Interaction l
Management Culture Manufacturing Defects 7s
4 PRA IMPLEMENTATION PLAN -
OVERALL OBJECTNES AND SCOPE Encompasses activities in NRR, RES, AEOD, NMSS, and Regions l
Informs Commission of staff progress via quarterly updates and briefings 8
l
l PRA IMPLEMENTATION PLAN Develop SRPs for risk-informed regulation Pilot applications for risk-informed regulatory initiatives i
Inspections Operatorlicensing Event assessment Evaluate use of PRA in resolution of generic issues Regulatory effectiveness evaluation Advanced reactor reviews Accident management Evaluating IPE insights to determine necessary l
follow-up activities
~
9
s RISK-INFORMED REGULATORY GUIDES
~
AND SRPS
~
DG-1061 - General SRP Chapter 19 - General Guidance to licensees guidance to staff l!
DG-1062-Application-SRP Section 3.9.7 -
specific guidance on application-specific i
in-service testing guidance on IST 4
DG-1063 - Application-SRP Section 3.9.8 - application-specific guidance on specific guidance on ISI-in-service inspection (ISI)-
under development under development DG-1064 - Application-Inspection guidance -
specific guidance on graded under development auality assurance
, DG-1065 - Application-SRP Section 16.1 - application-specific guidance on specific guidance onTS l
technical specifications 10
PRINCIPLES OF RISK-INFORMED REGULATION THAT GOVERN CHANGESTOTHE CURRENT LICENSING BASIS (CLB) i r
Maintain Defense-in-Depth r
r Meet current -
Maintain Sufficient Regulations Safety Margins V
7 integrated Decisionmaking implementation Proposed increases in risk and Monitoring and their cumulative effect Strategies Which are small and do not cause Address Uncertainties the NRC's Safety Goals to be exceeded a
1 11
~
l l
1 l
REGULATORY GUIDES AND SRPS (cont.)
i Principles of Risk-Informed Regulation for CLB changes j
The proposed change meets the current regulations.
This principle applies unless the proposed change is explicitly related to a requested exemption or rule change (i.e., a 50.12
" specific exemption" or a 2.802 " petition for rulemaking").
Defense-in-depth is maintained.
Sufficient safety margins are maintained.
Proposed increases in risk, and their cumulative effect, are small and do not cause the NRC Safety Goals to be exceeded.
Performance-based implementation and monitoring strategies are proposed that address uncertainties in analysis models and data and provide for timely feedback and corrective action.
12
i' REGULATORY GUIDES AND SRPS (cont.)
fll Expectations The overall risk management approach bses risk analysis to improve operational and engineering decisions broadly and not Just eliminates c
requirements the licensee sees as undesirable.
The acceptability of proposed changes should be evaluated
.by the licensee in an integrated fashion that ensures that all principles are met.
Core damage frequency (CDFl and large early release frequency (LERF) can be used as suitable metrics for making risk-informed regulatory decisions.
Increases in estimated CDF and LERF resulting from proposed CLB changes will be limited to small increments.
The scope and quality.of the engineering analyses (including l
l traditional and probabilistic analyses) conducted to justify the proposed CLB change should be appropriate for the nature and scope of the change and should be based on the as-built and as-operated and maintained plant.
13
i REGULATORY GUIDES AND SRPS (cont)
Expectations (cont.)
Appropriate consideration of uncertainty is given'in analyses and interpretation of findings.
The plant-specific PRA supporting licensee proposals has been subjected to quality controls such as an independent.
peer review.
Data, methods, and assessment criteria used to support regulatory decisionmaking must be scrutable and available for public review.
14
PRINCIPAL ELEMENTS OF RISK-INFORMED PLANT-SPECIFIC DECISION MAKING s
Traditional i
m m
PRA-Analysis 7
l1
-s t
I,'
l
's l
I
,s'
's i
llf
's l
's,'
I Pe..;....' -
Nfine
~
Submit
. Def.ine Engineering Implementation /
PropsM Monitonng Change Analysis Change Program Element 1 Element 2 Element 3 Element 4 i
e e
15
l REGULATORY GUIDES AND SRPS (cont.)
i Acceptance Guidelines Defense-in-depth is maintained l
a reasonable balance prevention of core damage, j
prevention of containment failure, and consequence mitigation is preserved over-reliance on programmatic activities to compensate j
for weaknesses in plant design is avoided system redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences of challenges ~to the system l
(e.g., no risk outliers) 1 defenses against potential common cause failures are preserved and the potential for introduction of new common i
cause failure mechanisms is assessed independence of barriers is not degraded defenses against human errors are preserved 16
REGULATORY GUIDES AND SRPS (cont.)
Acceptance Guidelines (cont.)
e Sufficient safety margins are maintained codes and standards or alternatives approved for use by the NRC are met safety analysis acceptance criteria in the current licensing basis (e.g., FSAR, supporting analyses) are met, or proposed revisions provide sufficient margin to account for analysis and data uncertainty 17
Increase in CDF a
a O
C I
I O
E.
f
.b*p
'g
.'f"$%. f))
i g-Z 05$
t's
~
O N
Y 3
%S "j;y yw 3
wt.
A g
p o
-w Dcp W
g M
m El 3
?$
.h h
g E g M" e g
- g
%[
p3 m
3 xe a
a 3
m
- n;-
'$ IE 3
- 1.
IL hp M.
E 2
,[N.,!
O o
2 m
- (4 l;f eO o
y a1 2
9 I
cs'h p Q
9"jt D
m'"1 g
g e
Se c
2[.r.
g W
$i S
r.!}
M W
Q o.
- 3 B
Mk:{l ma j!djA;23ip d
8 j;i Z
sQ gC
't I}f WW:
O O W s;
e s e=esm -
to g
S
-4 m3 e
aN q[.r M
- n-i
'T1 o
& b; r-s}.r r_-
m s
4o it m
=O s
y:
Q h
amm a w
b
$C
....,~,.
- .~.~.n,vm 4
FN#r.f$
g
- a. +~ # e =.e - m r
= e s.o
-O
@ c W
G o
O M
l!l lj1
,:l1,Il;lI1liI, t
llllI
- i1 I
c0Lemee.c LML L
L E
1 1
M 0
0 7
6 e
Sx n
s a
j M
N i
5 m
n o
T.
r s @.. c e
r e
~=- n f
a m
$n T w.
e
~ m L
+
r n
S a
d
.?
t l
Aa a
a s
7 L
f R
s
+.
n;s cr f
w i
a s
R
- fa t&.
r k
cg e
g
- m W_
M u
a v
i
+
tv n
e e e a
i
' n n
e f
as o
=
E wr m
Q'm h.
l-a ag
~
s p
E i
a a
R e
@m m
+
. -n t
r e
w..
a a a d
ly 1
iwk/,,;iger r
d j,l i
w__
m.. s N
nl 0
R u
O cy
-6'r9n%1q)*%>
e 1
{ -
t i
9 i
t E
T
~
eR s
e m
a n
A
~
s s
c
~r h
L Ge e
o
. i L
l F
O ue
.z in r
i e1 s
W da i
m x, i. L s ut *r;1ic.3 q0 t
u w
E s
s..
e
.k D
e e
"l. 8u. 4.{I"mn r!r 2;:v l
5
<t n
i c
nF y
er
. (
L s e E
q R
u F
e
)
n c
y i,,
REGULATORY GUIDES AND SRPS (cont.)
Acceptance Guidelines (cont.)
For a plant with a mean core damage frequency at or above 1 E-4 per reactor year (the Commission's subsidiary core damage frequency objective) or with a mean LERF at or above 1 E-5 per reactor year, it is expected that applications will result in a net decrease in risk or be risk neutral.
For a plant with a mean core damage frequency of less than 1E-4 per reactor year, applications will be considered which, when combined with the LERF guidelines described below:
Result in a net decrease in CDF or are CDF-neutral; or Result in increases in calculated CDF that are very small (e.g., CDF increase of less than 1 E-6 per reactor year); or
i j
REGULATORY GUIDES AND SRPS (cont.)
Acceptance Guidelines (cont.)
Result in an increase in calculated CDF in the range of 1E-6 to 1E-5 per reactor year, subject to increased NRC technical and management review and
- onsidering the following factors:.
The scope, quality, and robustness of the analysis (including, but not limited to, the PRA), including consideration and quantification of uncertainties; The base CDF and LERF of the plant; The cumulative impact of previous changes (the licensee's risk management approach);
Consideration of the Safety Goal screening criteria in the staff's Regulatory Analysis Guidelines, which define.what changes in CDF and containment performance would be needed to consider potential backfits; The impact of the proposed change on operational complexity, burden on the operating staff, and overall safety practices, and Plant-specific performance and other factors, including, for example, siting factors, inspection findings, performance indicators, and operational events.
A plant with a mean CDF in the range of 1E-5 to 1E-4 and an increase in CDF of up to 1E-5 will be subject to increased NRC technical and management review, as described above.
AND
~
21
l REGULATORY GUIDES AND SHPS (cont.)
Acceptance Guidelines (cont.)
For a plant with a mean LERF of between 1E-6 and 1E-5 per reactor year:
3 Result in a net decrease in LERF or are LERF-neutral; or
. Result in an increase in calculated LERF of up to 1E-6 per reactor year, subject to increased NRC technical and l
management review, as described above;.
t OR For a plant with a mean LERF of less than 1E-6 per reactor year:
Result in a net decrease in LERF or are LERF-neutral; or Result in increases in calculated LERF that are very small (e.g., LERF increase of less than 1 E-7 per reactor year); or Result in an increase in calculated LERF of up to 1E-6 per reactor year, subject to increased NRC technical and management review, as described above.
D
l STAFF AND MANAGEMENT R.ESPONSIBlLITIES e
Ensure that licensing submittals are identified and processed in accordance with risk-informed guidance; Identify current requirements that could be significantly enhanced with a risk-informed and/or performance-based approach e
Ensure the objectives of risk informed regulation are met:
Enhance safety decisions Efficient use of NRC resources Reduced industry burden Ensure adequate staff training on the use of risk-informed e
guidance and underlying PRA technical disciplines
_-.._.________.____._._m_____
m_____
b t
STAFF AND MANAGEMENT RESPONSIBILITIES NRR Projects Initial review of submittals Determine whether previously reviewed (consult with primary review branch-PRB)
In general, unless a previous precedent exists, review by PRB required Forward to PRB for action; copy to SPSB for information For particularly complex issues and first-of-a-kind reviews, DRP and ADT should meet to consider establishing a review team for the project
DECISION LOGIC FOR SUBMITTAL REVIEWS Changes to the Current Licensing Basis Staff Proposes increased Requirements - Use 50.109 Backfit Rule
.g.
(Regulatory Anahsis GuideNnes),.
-l A
Licensee Mair*E ~
~
" CURRENT LICENSING BASIS" '~~
Change Consistent
.with 50.59 Process
~
Licensee Requests Change Licensa Requests Change inRequirementsvia i in Requirernents Beyond
' Approved Staff Pamirians-:
Approved Staff Positions -
- (10 CFR 50.9(M12) f (10 CFR 50.90 92) s V
Licensee Requests Does Present.
Does Not Pmsent Change Consistent with.
Riskinfonnation '
- ApprovedMPosidon !
Riskinfonnation z
"Use Risk-informed hM
RGISRP" "NonnalStaff Review" 25
STAFF AND MANAGEMENT RESPONSIBILITIES e
Primary Review Branch Review application and determine level of PRA
~
support needed; considering:
Complexity Scope Risk Impact (perceived or calculated)
Relationship to previous reviews Existence of SRP/RG l
Need to rely ors general RG for guidance Define necessary support reviews Integrate SER inputs and issue to PM
i t
STAFF AND MANAGEMENT 4
ll RESPONSIBILITIES i
PRA and Severe Accident Branch Support i
f Provide support to PRB for applications which make l
use of probabilistic risk analysis as part of the license j
amendment justification Provide guidance as necessary for applying the general RG to new applications j
Consult with SRXB or Analytical Support Group as j
needed for success criteria i
~
j n
i
STAFF AND MANAGEMENT RESPONSIBILITIES c
Introductory Staff Training RG and SRP Initial orientation (this seminar) transmit guidance to staff with plans and expectations issue Office Letter for risk-informed licensing reviews PRA for Regulatory Applications (P-105)
PRA for Technical Managers (P-107) e Additional Training (under development)
Inspection Activities 2 week course (P-111: PRA Tech. & Reg. Perspectives) emphasis on practicality by using case studies cover inspection activities of Region and NRR cover revised inspection guidance on use of PRA Other in-depth PRA technical training e
NUREG/BR-0228, Level 1,2,3
i i
i PRAiMPLEMENTATION PLAN KEY NEAR-TERM ACTIVITIES e
issue Regulatory Guides and Standard Review Plans (RES/NRR)
Develop New Application-Specific Guidance for l
Steam Generator Performance (NRR/RES)
Complete PRA Pilot Reviews (NRR) issue Risk-informed Inspection Guidance (NRR) e Train Staff on Use of Risk-informed Guidance (NRR/AEOD)
Address Remaining issues in Commission SRMs l
l (RES/NRR/AEOD) as l
i i
PILOT APPLICATIONS OF PRA GQA (South Texas, Palo Verde, Grand Gulfi e
Risk-informed IST (Palo Verde, Comanche Peak)
Risk-informed ISI (Surry, ANO, FitzPatrick, Vermont Yankee)
Risk-informed Tech Specs (CEOG) e
~
l
~
so
l!
OBJECTIVES AND EXPECTATIONS FOR USE OF PRA INSIGHTS IN INSPECTION ACTIVITIES t
I
.i Objectives:
i.
Integration of Risk information with other non-PRA factors help to Plan inspection Activities Evaluate the Significance of Inspection Observations and Findings
~
Evaluate the Adequacy of Licensee Proposals or Programs which have Probabilistic Risk Elements iiiTheir Bases
~
4 i
31 1
~
OBJECTIVES AND EXPECTATIONS FOR USE OF i
j PRA INSIGHTS IN INSPECTION ACTIVITIES (cont.)
Expectations.
Use of PRA Complements Traditional Engineering Analysis Consistency in the Use and Communication of Risk Related information within the inspection Program t
~
PRA results do not provide the sole basis for imposing Regulatory Requirements.
Quar.litative thresholds alone do not determine acceptability i
PRA calculations can potentially result in a Larp Resour.ce impact to Licensees inspection Activities should focus on the Dominant Accident e
i Sequences and Success Criteria
- ' Utilize PRA information to Characterize the Significance of t
inspection issues.
l m
i l
ISSUE RISK-INFORMED i
INSPECTION GUIDANCE l
IMC 2515 Appendix C (complete) 50.59 Evaluations (12/97)
Regular Maintenance Observations (12/97) i GQAinspection Procedure (draft)I:12/97) l GQA Inspection Procedure (final) < 3/98) l 1
i i
i M
~
,