ML20198E862
| ML20198E862 | |
| Person / Time | |
|---|---|
| Issue date: | 12/01/1998 |
| From: | Isom J NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9812240162 | |
| Download: ML20198E862 (93) | |
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w Dacember 1, 1998 MEMORANDUM TO: File FROM: James A. Isom, Operations Engineer, O ~
Inspection Program Branch /
l Office of Nuclear Reactor Regulatiori
SUBJECT:
SUMMARY
OF THE NOVEMBER 24,1998 MEETINGS WITH THE NUCLEAR ENERGY INSTITUTE TO DISCUSS OPTIONS FOR REVISING THE REGULATORY OVERSIGHT PROCESS l On November 24,1998, a public meeting was conducted between NRC and NEl to continue
- exchange of information and views in developing improvements to NRC processes for overseeing performance of operating reactors. The meeting agenda and written information exchanged are attached.
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gng h.w l yozyg CONTACT: James A. Isom 301-415-1109
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Attachments: 1. Agenda-
- 2. List of Attendees
- 3. Performance Indicator Table for Selected Plants
- 4. Interim Guidance on Implementation of Risk-Informed Review Processes
- 5. NRC Power Reactor Baseline inspection Program DISTRIBUTION:
PUBLIC N
PIPB R/F F. Gillespie M. Johnson C. Holden A. Madison J. Isom T. Frye J.Jacobson D. Gamberoni R. Barrett .
P. Baranowsky B. Mallet DOCUMENT NAME: MTG1124. SUM To receive a copy of this document, indicate in the box: "C" = Copy without enclosures "E" = Copy with enclosures "N" = No copy OFFICE PIPB: DISP _ l PlP8: DISP h _- l l l NAME- JAlsomjfM MRJohnso'nlNX DATE 12/ O 7 /98 / - 12/ ) /98 ( )
OFFICIAL RECORD COPY
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! November 23,1998 NRC/NElPUBLIC MEETING L
l 9:00AM Welcomelintroduction Alan Madison l
l l 9:10am Assessment Program Development Michael Johnson 9:25am inspection Program Rebaselining Bruce Mallet Discussion of Preliminary Draft 10:30am Framework Development Patrick Baranowsky l
11:40am NElFeedback Tom Houghton l
12:00 noon Adjourn l
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NOVEMBER 24,1998, NRC/NEl MEETING ON PERFORMANCE ASSESSMENT PROCESS IMPROVEMENTS i
l SIGN-IN SHEET l
NAME ORGANIZATION PHONE NUMBER R. J. Acosta Florida Power and Light 561-694-3656 P. W. Baranowsky NRC/AEOD 301-415-1111 Stephen D. Floyd NEl 202-739-8078 David Garchow PSEG 609-339-3250 Don Hickman NRC 301-415-6829 Tom Houghton NEl 203-739-8107 Jeff Jacobson NRC/NRR 301-415-2977 Michael Johnson NRC/NRR 301-415-1241 Steve Lockfort NYPA 914-681-6868 Alan Madison NRC/NRR 301-415-6412 l Bruce Mallet NRC/R-Il 301-415-1425 Jim McCarthy Virginia Power 804-273-2699 G. W. Parry NRC/NRR 301-415-1464 l Jeff Reinhart INPO 770-644-8217 W. H. Rutand NRC/R-l 301-415-1380 l
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l Attachment 2 i
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RCS Safety Sefety Safety System Perforumance Indweser(Unevailetility) .
Perferumence Pinot Coolant RCS Centsament thsplanned Systems Systeam Tements Cherneterimetism Activity ' trakage' leakage' Screens @ p,gg,g > 20% ' g AC' Insecteam' AFW 8 (B%lts)
Escutmat(NEI) Summer G G No Data G(1) Gil) G(1) GG) G(0 0ll) G(0.012) G (0.007) NA i NA e Turkey Pt 3 G G G G(1) O (!) G(3) G P) G(0.007) G (0.009) G (0.018) i Twkey Pt 4 G G G G G) G(1) G G) W(12) G (0.007) G (G.011) G (0.013) NA l i Seperior(NRC) Callaway G(3) G(0) G(3) G(5) W (0.021) G (0.012) G (0.013) NA Vogtle I G(2) G Q) G(3) G (3) G(0 006) G (0.004) G (0.002) NA t Vostie 2 G G) G(1) GG) G(3) G (0.006) G(0.003) G (0.006) NA 2 Average CalvertC 1 G G W(1Q) G(3) G(3) G(3) G(6) W (0.027) G (0.013) G (0.012) NA (NEI)
Calvert C 2 G G W(1Q) W (4) G(3) G G) G(9) W (0.027) G (0.00s) W (0.024) NA Average Davis-Besse G(I) G(0) G C) G(5) G (0.004) G(0 008) G (0.005) NA (NP.C) ,
Pt Besch l G G) G(1) W(12) G(3) W (0.051) G (0.003) O (0.013) NA ,
Pt Beach 2 G(1) G(2) W(10) G(3) W(0.051) G (0.003) G(0.0ll) NA TMil U(l) G(1) G(2) G(4) W (0.021) Wi0.018) G (0.004) NA Cooper G(1) G(1) W(ll) G Q) G (0.0ll) G(0.029) NA G (0.014) flope Creek W(5) G(2) W0) G G) G (0.018) W(0.042) NA G (0.006)
Declunseg St Lucie l G G G W(5) W(4) G(3) W(14) G (0.016) W(0.027) G( 0.010) NA t I
Sa lacie 2 G W(lQ) G G G) G(1) G(1) W(22) G (.0.016) G(C.Olo) G (0.019) NA Ft Calhoun W(5Q) W(1Q) W(IQ) G(3) G Q) W (5) G(4) G ( 0.012) G (0.001) G (0.003) NA R(lQ) 4 r
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Sarety Systems Perfemnace Indicator (Unavailabihty)
Safety Safety R(3 Transients RCS Centaimist Uaplanned SystennS Systeen ,
Plant Ceolant > 20% s Perforumance ArtMey ' ?sabage' Isaksee' Scrsess'# Ap p,";g,,,,,a J Charactertraties AC* tnjecnem' AFW' (BWRs)
G (0.0t9) G (0.015) NA G (0.0t?)
No data G(1) G(2) W(9) G(8)
C..l 2 No data G Watch list G(0019) G (0.018) NA G(0.01I)
(NEI) Nodata W(4) G(2) W (9) G(7)
C..Aa 3 No data G G(0.009) G (0.009) NA G(3) W (II) W C6) G(0 014)
Salem i G G Dad rM W(5)
G(0.010) G (0.005) NA GQ) W(!!) W (23) G(0.014)
G G Ba ipict? G Q)
Salem 2 W (0.026) G (0.071) NA G (0.006)
No data G(t) G(0) W (8) W(13)
Quad Cities 1 G W(10) i NA G (0.001) j W(14) W (0.026) W (0.037)
G No date W(5) G fI) W(l3)
Quad Cities 2 G G(0 012) W(0.021) NA G (0.009)
G(3) W(8) G (7) 1 Me i G G No data G(3)
G (0.029) NA G (0.008)
G(3) W (6) G (4) G(0.012)
G G No data W (4) e IMalle 2 G (0.003) G (0.006) NA Datchlist G(9) W (0.026) ladian Pt 3 G(3)
(NRC) NA G(t) G(0.009) G (0_012) G (0.007)
G(I) G (2) W (9)
Crystal R 3 I Performance assessed using thresholds proposed by NEl. Where indicator has entered the white or red region, the number of quarters w 2 Values repotted in () are the maximum number of events occurnng within a calendar year over the period 1993 to 1997; perform 3
Values reported in () are the maximum number of events occurring within a calendar 3 ear over the period 1993 to 1997; performan 4 Performance assessed using a threshold of 0.02 5 Performance assessed using a threshold of 0.015 for PWRs and 0.03 for BWRs 6 Performance assessed using a threshold of 0.02 7 Perfonnance assessed using a threshold of 0.015
ocarg p & UNITED STATES j
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NUCLEAR REGULATORY COMMISSION t WASHINGTON, D.C. 20665 0001 [L
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- p$ October 30, 1998
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MEMORANDUM TO: NRR Staff and Management FROM: Gary M. Holahan, Director W Division of Systems Safety and Analysis Office of Nuclear Reactor Regulation
SUBJECT:
INTERIM GUIDANCE ON IMPLEMENTATION OF RISK-INFORMED REVIEW PROCESSES The purpose of this document is to assist the NRR staff in using risk-informed approaches in various regulatory activities.
One of the top priorities in the Chairman's Tasking Memorandum is for the Nuclear Regulatory Commission'(NRC) to become more risk-informed. As part of that process, we have developed the attached interim guidance on how the Office of Nuclear Reactor Regulation (NRR) should go about performing reviews of risk-informed licensing actions or activities. This information complements guidance in Standard Review Plan Section 19. Because there are an extensive number of risk-informed policy ma'ders currently being considered in parallel (e.g., risk-informed modification of Part 50, pilot risk-informed reviews,10 CFR 50.59 changes), this guidance is interim, with final guidance expected in mid 1999.
The Probabilistic Safety Assessment Branch is available to help individual branches implement this guidance. Feedback on the risk-informed implementation process should be provided to the contact below.
Attachment:
' As stated CONTACT: Glenn Kelly, SPSB/DSSA (301)415-1075 l
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ATTACHMENT INTERIM GUIDANCE FOR IMPLEMENTING RISK-INFORMED REVIEWS l WHAT IS RISK-lNFORMED REGULATION?
Risk-informed regulation is the use of insights and results derived from Probabilistic Risk l Assessments (PRAs) in combination with traditional engineering (deterministic) analyses to focus licensee and regulatory attention on issues commensurate with their importance to safety.
It is the Commission's desire that the Nuclear Regulatory Commission (NRC) and the industry i make appropriate use of risk-informed regulation in their daily work. The objectives of risk-informed regulation are to enhance safety decisions and make more efficient use of industry and NRC resources.
In its 1995 Policy Statement on the use of PRA, the Commission determined that use of PRA technology should be increased in all regulatory matters to the extent supported by state-of-the-art PRA methods and data. This use should complement the NRC's traditional engineering (deterministic) approach and support defense-in-depth. PRA and associated analyses should be used in regulatory matters to reduce unnecessary regulatory burden and conservatism.
l Note that the intent of this policy is that existing rules and regulations cannot be ignored simply because they appear to have low risk significance. Absent a rule change or exemption, regulations must be cdmplied with.
WHY DO WE NEED GUIDANCE ON RISK-INFORMED REVIEWS AND DECISION-MAKING 7 We need guidance primarily because the Nuclear Reactor Regulation (NRR) staff and managers are unclear about how risk informed regulation is to work in their particular area within the agency. Staff have heard about risk-informed regulation, but many don't know what it is and don't know how to go about implementing it.
I DEFINITIONS AND GUIDANCE Definitions risk-informed licensing action or activity Any licensing action or activity that uses quantitative or qualitative risk assessment insights or techniques to provide a key comoonent of the basis for the acceptability or unacceptability of the proposed action. Mere mention of quantitative or qualitative risk insights does not in itself make a licensing action or activity risk-informed.
Risk Informed Licensing Panel (RILP)
The Risk informed Licensing Panelis made up of Division Directors in NRR that participate in licensing reviews, as well as representatives from the Office of General Counsel (OGC) and the Office of Nuclear Regulatory Research (RES). One of the main purposes of the panelis to streamline the review of risk-informed licensing actions by serving as a focal point for resolution l
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2 ATTACHMENT of technicalissues and for guidance on policy implementation to the NRR staff. This panel will-provide a forum for the staff, licensees, owners groups, and the public to receive management attention on risk-informed issues. The panel will also monitor the overall implementation of risk-
! informed licensing actions.
verv low risk sionificance i
l An issue in which risk is expressed numerically is of very low risk significance if it results in a
! risk decrease, is risk neutral (i.e., it has no effect on risk or the change is too small to measure i
i accurately), or results in an increase of less than -1E-6 per reactor year (mean value) to core damage frequency (CDF) estimates, or an increase in large early release frequency (LERF) of less than -1E-7 per reactor year.
fow to moderate risk sianificance I
! An issue in which risk is expressed numerically is of low risk significance if it results in an increase to CDF estimates in the range of -1E-6 to -1E-5 per reactor year (mean value) or an l increase in LERF in the range of ~1E-7 to -1E-6 per reactor year.
substantial risk sianificance An issue in which risk is expressed numerically is of substantial risk significance if it results in an' increase to CDF estimates greater than -1E-5 per reactor year (mean value) or an increase in LERF greater than -1E-6 per reactor year. Note that a " substantial risk increase" should not normally be approved. In fact, approving a change that allows such a risk imrease would result in a risk contribution meeting the criteria for consideration of a backfit analysis and possible action to correct the very situation. ' Guidance that would allow such a circular " approval" and l " consideration for backfit" cycle would be inappropriate.
Ruldaage on the Review Process for a Risk-Informed Submittal l Project managers (PMs) should send all risk-informed submittals to the Probabilistic Safety i
Assessment Branch (SPSB), NRR, for review or concurrence. In the near term, SPSB is expected to be directly involved in all risk-informed licensing actions.
The project manager should (1) Determine if the submittalis risk-informed (using the definition above).
(2) Inform Randy Hall, the lead PM for Risk-Informed Licensing Actions (RI-LAs) that the PM l has received an RI-LA.
(3) Assign a TAC number and mark it as risk-informed (i.e., the review method should be "TSR" or "PMR.")
(4) Identify a lead review branch, with SPSB marked for PRA review support.
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3 ATTACHMENT (5) Discuss the scope of the review required with the responsible technical branches.
(6) Send a copy of risk-informed submittals to SPSB.
(7) Bring conflicts between branches, divisions, or offices regarding risk-informed licensing submittals to the Risk-Informed Licensing Panel (RILP) by contacting Sam Lee (SPSB),
Richard Barrett (SPSB), or Gary Holahan (DSSA).
(8) Ensure that requests for additionalinformation (RAls) are focused and are seeking a scope and depth of information in line with the risk significance of the licensing action or activity.
(9) Send a filled out Risk-informed License Amendment Cumulative Risk Tracking Form (see Page 19-17 of Standard Review Plan Section 19 or updated form if available) when the licensing action or activity is complete and signed off on by the PM or project director.
The lead review branch chief should (1) Ensure that RAls dealing with a risk-informed submittal are sent to SPSB for review and concurrence. Note that review of RAls is frequently delegated to section chiefs.
(2) Bring conflicts between branches, divisions, or offices regarding risk-informed licensing submittals to the RlLP by contacting Sam Lee (SPSB), Richard Barrett (SPSB), or Gary Holahan (DSSA).
The NRR lead branch reviewer should (1) Coordinate or consult with SPSB regarding determination of the risk significance of the issue (i.e., does it cause a decline in overall risk; is it risk neutral; or does it cause an increase in overall risk, and if so, how much?).
(2) Work with SPSB to identify strengths and limitations of a licensee's risk evaluation.
(3) Follow the guidance of Standard Review Plan Chapter 19. Note it is expected that the lead reviewer wil' use good judgement in developing a scope of review commensurate with the risk importance of the issue.
(4) Work with SPSB to determine an appropriate balance between traditional engineering (deterministic) and probabilistic review, based on the risk significance of the licensee's submittal request.
l l (5) Assess all non-risk-informed licensing action and activity submittals to seek to identify if there are any unaddressed, potentially significant risk effects (e.g., potentially significant changes in CDF, LERF, design margins, or defense-in-depth) that approval of the licensing action or activity could precipitate. If the reviewer suspects that there is such a potential, the nature of the concern should be documented and forwarded along with the submittal to SPSB for joint review and consultation.
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The SPSB branch chief should l
(1) Be responsible for the timeliness of the SPSB review of the risk-informed submittal (2) Ensure that RAls generated by SPSB are appropriate for the risk significance of the issue.
(3) Concur in the appropriate level of traditional engineering (deterministic) and probabilistic i- review. This function is normally delegated to section chiefs.
(4) Bring conflicts between branches, divisions, or offices regarding risk-informed licensing l submittals to the RILP by contacting Sam Lee (SPSB), Richard Barrett (SPSB), or Gary Holahan (DSSA).
(5) Ensure the Risk-Informed License Amendment Cumulative Risk Tracking Form is correct and is attached to the safety evaluation report.
The SPSB reviewer should l (1) Help the lead reviewer determine the risk significance of the risk-informed submittal.
l (2) Help the lead reviewer determine the level of traditional engineering (deterministic) and risk review needed for the submittal, based on the risk significance of the issues involved.
l (3) Provide a detailed review or audit of the risk-informed submittal, including a description of the depth and scope of the review performed.
l (4) Assist project managers, as necessary, to complete the Risk-informed License Amendment Cumulative Risk Tracking Form.
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NRC POWER REACTOR t
BASELINE INSPECTION PROGRAM NOTE: THIS IS A PRELIMINARY DRAFT IN PROCESS. NOT FOR DISTRIBUTION.
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NRC POWER REACTER BASELINE INSPECTION PROGRAM ;
- l. Program overview II. PART 1 of Program: Inspectable Areas I
A. Inspectable areas required in the basehne program '
List by cornerstone Linked to Agency mission B. Basis documents for inspectable areas Scope of inspectable area Linkage of inspectable area to performance indicators Justification for why each inspectable area is in baseline program Activity and SSC selection methodology i
C. Process for selecting activities to be inspected within inspectable areas Generic - first cut approach using risk chart Plant specific - using SRA input / licensee site specific data l.
D. Process for determining frequency, how many activities and how much I time to expend inspecting within each inspectable Area L
l Ill. PART 2 of Program: Process for verification of performance indicators IV. PART 3 of Program: Process for evaluating problem identification and resolution l
programs.
V. Inspection Procedures 4
VI. Interface with other NRC processes Allegations Significant event followup for AITs and llTs
! Assessment l Enforcement Training Vll. Risk information matrices (RIM)
Vill. Definitions of terms used in the basefine inspection program IX. Projected resources for baseline inspection program
- X. Program Feedback and Assessment i
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Q Section 1: Program Overview / Concept of Operations .
The NRC power reactor baseline inspection program defines the planned activities to evaluate licensee performance over a 12 month period. The overall objective of the program is to monitor all power reactor licensees at a defined level of effort to assure the licensees are using l- radioactive materials safely as defined by the goals specified in the NRC's Strategic Plan.
L i The baseline reactor inspection program is divided into three parts: ,
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[ Part 1: Inspection in each inspectable area.
L Part 2: Inspection to verify established performance indicators.
- l. Part 3: Inspection to review licensee programs for identification and resolution of probletns -
i and/or issues, j
j The program concept for Part 1 is that all areas where there is a need to inspect a licensee's L performance are defined first. These are listed as "inspectable areas" in Table 1 of the document. Next, the baseline program is defined as only requiring inspection in those inspectable areas that are not bounded by performance indicators or in areas where the l performance indicators are limited in coverage. The inspectable areas that are Dat required because of sufficient coverage by a performance indicator will be marked in Table 1 by an asterisk. These areas will also be clearly explained in the basis document for each inspectable area.
l - Another key concept for Part 1 is that each inspectable area will have a basis document, which p
describes the scope of the inspectable area and the reasons why the area is included in the I baseline program. Reasons include (1) the linkage to the NRC's mission, (2) the linkage to the key attribute of a comerstone of safety that the inspectable area is measuring, and (3) the risk information that was used to explain why there is a need to include the inspectable area in the baseline program. Note that risk information is defined as that from risk analyses, from history of problems (i.e., experience) and from deterministic / defense-in-depth considerations. A third key concept in Part 1 is that the Regional managers and inspectors will plan the type and nurr,ber of activities to to inspect each year for each reactor site, based on the guidance contained in the selection methodology and frequency sections of the basis documents. This guidance was based on risk information derived into a Risk Information Matrix (RIM).
Use of risk information is factored into this part of the program in four ways: (1) inspectable areas are based on the importance as a measure of an attribute into one of the seven
. comerstones of safety, which lead to our NRC mission. The comerstones are: initiating events, l mitigation systems, barrier integrity, emergency preparedness, occupational dose, public dose and physical security, (2) the frequency, how many activities, how much time to inspect activities in each inspectable area is based on risk information in a Risk Information Matrix (RIM).
Information in the RIM is explained in Section VI of the baseline inspection program, (3) selection of activities to inspect in each inspectable area is based on the use of c risk chart, modified by plant specific infortlation, and (4) inspectors are trained in use of risk information.
The concept for Part 2 of the p: 9 ram is that there will be verification of performance indicators used in lieu of inspection. This verification will be at each reactor site and across reactor sites.
t The concept for Part 3 of the program is that there will be inspection of licensee programs for
' identification and resolution of problems and/or issues in conjunction with inspection of activities l
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'in other inspectable areas. There will also be a defined overview inspection of the licensee's overall program in this area each year.
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Section 11: PART 1 of Program: Inspectable Areas This Part of the inspection program consists of the annual review of areas to inspect (i.e.,
inspectable areas) at each power reactor facility. The inspectable areas and their link to the attributes they are measuring is depicted in Chart 1 attached. The inspectable areas are also listed in the attached Table 1 by comerstone.
A. Inspectable Areas required in the baseline program.
There are specified inspectable areas that will not currently be measured as part of the baseline inspectin program because they are sufficiently covered by review of performance indicators submitted by licensees. These inspectable areas will be indicated in Table 1 by an asterisk next to the area.
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' Table 1: Inspectable Areas by Cornerstone i
l A. Initiating Events Comerstone Inspectable Areas (Important attributes in parentheses)-
L Adverse weather preparations (extemal factors) l Equipment alignment (equip. perf and config. control) l Emergent work (configuation control)
L Fire protection (extemal factors)
Flood protection measures (extemal factors)
General plant walkdowns identification and resolution of problems / issues '
l Inservice inspection activities (equipment performance)
L
. Piping system erosion and corrosion (equipment performance)
Pre-job briefs (human performance)
B. Mitigation Systems Comerstone inspectable Areas (Important attributes in parentheses):
' Equipment alignment (equip. perf. and config. control)
Event followup (equipment performance)
Fire protection (design and equip. perf.)
Flood protection measures (design)
General plant walkdowns Heat exchanger performance (equipment performance)
Identification and resolution of problems / issues inservice testing - Section XI Licensed operator requalification (human performance)
Maintenance rule implementation (equipment performance)
Maintenance work prioritization control (configuration control)
Off-normal plant operations (human performance)
Operability evaluations (equipment performance)
Operator workarounds (design)
Permanent plant modifications (design)
Post maintenance testing (equipment performance)
Refueling activities (configuration control)
Safety evaluations (design)
Safety system design and performance capability (design and ' equip. perf.)
Technical specification surveillance testing (equipment performance)
Temporary plant modifications (design and config. control)
!' ' C. Barrier Integrity Comerstone l
Inspectable Areas (Important attributes in parentheses):
Event followup (barrier and equipment performance)
Fuel barrier performance (cladding performance and config. control)
' General plant walkdowns identification and resolution of problems / issues i
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- o . li Inservice inspection activities (barrier performance)
Large containment isolation status verification (SSC barrier performance)
Licensed operator requalification (human performance)
Maintenance work prioritization control (configuration control)
Permanent plant modifications (design) l . Post modifications testing (design)
Refueling activities (fuel: human perf., design and config. control)
D.: Emergency Preparedness Comerstone
! Inspectable Areas (Important attributes in parentheses):
L Alert and notification system availability (equipment readiness)
Drill / exercise performance (ERO performance)
Emergency action level changes (procedure quality)
Emergency response organization augmentation (ERO readiness)
- identification and resolution of problems / issues Ucensed operator requalification (ERO performance)
E. Occupational Exposure Comerstone 1 I
Inspectable Areas (Important attributes in parentheses):
l Access control to radiologically significant areas (6quipment/ facilities) !
ALARA planning and controls (program / process)
L Event followup General plant walkdowns identification and resolution of problems / issues Radiation monitoring instrumentation (equipment)
Radiation worker performance (human performance)
F. Public Exposure Comerstone Inspectable Areas (Important attributes in parentheses):
. Event followup Gaseous and liquid effluent treatment systems (equipment)
Identification and resolution of problems / issues Radioactive material processing and shipping (program / process)
Radiological environmental monitoring program (program / process)
G. Physical Security Cornerstone Inspectable Areas (Important attributes in parentheses):
Access authorization program (personnel screening, FFD, behavior obs.)
Access control (search, ID and authorization, security locks and key control)
General plant walkdowns
- l. Identification and resolution of problems / issues j Physical protection system (barriers, intrusion detection, alarm assessment) l Security plan and procedures modifications (system design control)
Response to contingency events (protective and implementation strategy)
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B.' Basis documents forinspectable areas i Attached is a listing of basis documents for each inspectable area. They are listed alphabetically, i
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Insnectable Area P.agg Access Authorization Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 f
Acce s s Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................ ... . . .. 2 t
Access Control to Radiologically S!gnificant Areas . . . . . . . . . . . . . . . . . ........ .. 4 Adverse Weather Preparations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...... ... 6 ALARA Planning and Controls . . . . . . . . . . . . . . . . . . . . . . . . . . . . ........... ..8 Alert and Notification System Availability . . . . . . . . . . . . . . . ...... . ............ . .9 Drill and Exercise Performance (DEP) . .. . .. ...... . ..... .. ........ . 10 Emergency Action Level Changes . ........ . . ..... . . ..... . . . . . . . . . 12 Emergency Response Organization Augmentation . . . . . . . . .... . . ... .. . 13 Emergent Work . . .... .. .......... ....... . . ..... . .. ... .... .... 14 Equipment Alignment . . . . .. .... . . . . . .... . . . .... . .. .... 16 Event Followup . . . . . . . . . ..... ........... . .. . . ... . .. .... 17 t
.. Fire Protection . . ....... .. .. ...... .. .. ...... ...... .... ..... . . 18 Flood Protection Measures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .... 19 Fuel Barrier Performance . . . . . . . . . ...... .... ..... . .. . . .... .. .20 Gaseous and Liquid Effluent Treatment Systems . . .. .. .. .. . ... . .. . . . . . . 21 l
General Plant Walkdowns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . .. . 22 I Heat Exchanger Performance . . . . . . . . . . . . . . . . ... . ....... . ...... . 24 i
Identification and Resolution of Problems / issues ... .... .. . ..... . ....... .25 Inservice lnspection Activities . . . . . . . . . . . . . . . ... ..... . . .... .. . . . .. 2 7
. inservice Testing - Section XI . . . . . . . . . . . . . . . . . . . . ..... ..... . .... ...... 29 Large Containment isolation Valve Leak Rate and Status Verification . . . . . . . .. .. .. . 30 Licensed Operator Requalification . . . . . . . . ... ... . .... ... .. .. .. . . . 31 Maintenance Rule implementation . . . . . . . . . . . ... . .. .... . . ...... . . . 32 Maintenance Work Prioritization and Control . ...... . . .... . .. . .. . . 33
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Off-Normal Plant Operations . . . . . , . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . ....... . . 34 Operability Evaluations . . . . . . . . . . . . . . . . . . . . . . . ............ ...... ... .. . 35 Operator Work-Arounds . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 36 l
l Permanent Plant Modifications . . . . . . . . . . . . .............. ......... . . . . . . . . . 37 l
Physical Protection System (Barriers, Intrusion Detection System, and Alarm Assessment) . 38 Piping System Erosion / Corrosion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 39 Post Maintenance Testing . . . . . . . . . . . . ... .... .. .......... ............... 40 Pre-job Briefings . . . . . . . . . . . . . . . . . . . . . . . .... . ....... ..... ... . .. 41 Radiation Monitoring Instrumentation . . . . . . . . . . . , . . . . . . . . . . . . ...... . . 42 l Radiation Worker Performance . . . , ....... . ... .. ... ..... .... . 44 I
t Radioactive Material Processing and Shipping . . . . ... . . . . .. .... . .. .45 Radiological Environmental Monitoring Program (REMP) . . . . ........ ... .. . , , 47 Refueling Activities . . .. ........ ........... .. .. .. .. . ... .. .. 48 Response to Contingency Events (Protective Strategy and implementation of Protective Strategy)
. .. ..... .................. ........ . ... . . ... .. .. 49 Safety Evaluations (10 CFR 50.59) . . ........ ..... ... .. .. ........ . 51 Safety System Design and Performance Capability .. . ....... .. .. ... ..... . 52 l Security Plans and Procedures (Includes Security System Modifications) .... . .. ... . 53 Technical Specification Surveillance Testing . . . ..... ... .. . ... . . . . . 54 Temporary Plant Modifications . . . . . . . . . .... .... ... . .... .. .. .. . . . 55 f
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INSPECTABLE AREA Access Autn' orization Program (Personnel Screening, Fitness-for-duty and Continuing Behavior Observation)
Scope i
Verify that the licensee is properly implementing their personnel screening and fitness-for-duty l~ program, granting, denying, and revoking unescorted access authorization into the protected area, as appropriate. The areas to measure are the Personnel Screening Process, the Fitness-For-Duty (FFD) program and the Behavior Observation Program.
Basis l Inspection of this item supports the Physical Protection comerstone.
The licensee is required to maintain an access authorization program granting individuals unescorted access to protected and vital areas with the objective of providing high assurance that the individuals are trustworthy and reliable and do not constitute an unreasonable risk to L public health and safety including the potential to commit radiological sabotage.
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' This is a risk significant area because the personnel screening and fitness-for-duty processes are used to verify personnel trustworthiness prior to granting unescorted access to the site protected and vital areas. The establishment of trustworthiness for persons granted unescorted access to the protected area is a major component of protection against the insider threat of radiological sabotage. The fitness-for-duty behavioral observation process is used to monitor the continuation of trustworthiness for persons authorized unescorted access and escorted visitors.
An unauthorized individual with malevolent intent or an individual under the influence of drugs
! could be granted unescorted access due to human or program failure. The frequency of this type of event has been low but the safety significance of this type event is high. The probability of a single individual causing a radiological release is unknown although the consequences of an individual causing a radiological release release can be high depending on the individual's j knowledge of plant systems.
Historically, licensees have been effectively implementing the personnel screening and fitness-for-duty programs.
The performance indicator for this area will be based on the number of reportable events. This data is currently available and there are regulatory requiremtents to report significant events in the areas of Personnel Screening and Fitness-For-Duty. The Behavior Observation significant l events are captured in the FFD reporting requirements. However, a minimum baseline inspection should be conducted of the program process and human performance attributes.
1 l Selection Methodology l-The licensee's performance will be verified by reviewing audits and quarteriy/ semiannual data se.mitted to the NRC. Validation of performance would be conducted by inspection of program process and human performance attributes.
i-j - Frequency i
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! i l' l Baseline inspection of this item would be conducted every two years. RegionalInitiative and .
Reactive inspections would be conducted as needed. TWD 11/19/98 l .-
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-INSPECTABLE AFtEA- Access Control (Search of Personnel, Packagss, and Vehicles; identification and Authorization)
Scope Verify that the licensee has access controls and equipment in place designed to detect and prevent the introduction of contraband (firearms, explosives, incendiary devices) into the protected area that could be used to commit radiological sabotage and to assure that only authorized personnel are permitted unescorted access to the site protected area. The identification and Authorization process is to assure that, once personnel have been screened to verify their trustworthiness, those persons have a need for access and to confirm that only those persons who have be'en screened and have a need are granted access to the plant including vital areas. The equipment involved is metal detectors, explosive detectors, x-ray machines, computers, and keycards, Basis Inspection in this area supports the Physical Protection Comerstone.
' There is no performance indicator for this inspectable area. During discussions between NEl, industry, and the NRC, it was concluded that meaningful tracking data on the performance of those processes was not practical since much of the performance is dependent on the quality of the implementation of the tasks. Assessment of this area will be through the baseline inspection program.
The areas to measure are the effectiveness of the search function (personnel, packages and vehicles) and the identification and Authorization. The search function is to prevent the introduction of contraband (firearms, explosives, incendiary devices) that could be used to commit radiological sabotage. The search function for detection of firearms, explosives and incendiary devices on individuals, in packages, or vehicles, is accomplished by equipment or a hands-on search. The licensee must also positively control all points of personnel and vehicle access into vital areas.
The frequency of an unauthorized individual or the introduction of contraband into the protected or vital areas is low. The risk-significance consequence to radiological sabotage is low to moderate.
Selection Methodology The capabilities of access control equipment to perform their intended functions will be determined by requesting and observing the licensee conduct performance I:ased testing of the equipment using their testing methods and devices.
The processing of personnel, packages, and vehicles entering the protected area should be observed to determine if security officers can effectively perform all of the tasks assigned to them by procedures.
Verify that keys to vital areas are controlled and protected in accordance requirements.
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Inspection of this area should be conducted annually. The access control equipment is a high maintenance area. Each time an equipment failure is repaired it must be operability tested. In many instances the test to return the equipment to service is conducted the security officer on-i duty in the area at the time. Experience has shown that equipment may not always be l adequately performance tested.
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l lNSPECTABLE AREA: Access Control to Radiologically Significant Areas Scope i
This area will verify that the licensee has implemented offective Radiation Protection (RP) l barriers to prevent an uncontrolled access to an airbome, high (HRA) or very high (VHRA) j radiation area that could potentially result in an overexposure. An RP barrier includes:
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' identification and control of the hazard; procedures; physical barriers or engineered controls (e.g., barrier ropes, locked door, shielding or ventilation system); radiological surveys and monitoring (e.g., RP technician coverage, personnel alarming dosimeter or remote -
morntoring/ surveillance); and radworker training.
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Basis j inspection in this area supports the Occupational Dose comerstone.
l This inspection will review the licensee's performance in instituting the physical and l administrative controls defined in Subparts G, I and J of 10 CFR Part 20, applicable technical i specifications (TS), and licensee procedures for airborne areas, HRAs and VHRAs and worker adherence to these controls.
I The NRC position is that the radiological risk (i.e., exposure) to a worker be within the occupational exposure limits defined in 10 CFR Part 20 and ALARA to minimize the potential for health effects. Collectively, the aforementioned controls provide a " defense-in-depth" against a significant exposure from an uncontrolled access. Industry experience has identified frequent occurrences where the failure of multiple barriers resulted in an uncontrolled entry and, in some l cases, a significant exposure.
The established Performance Indicator (PI) does not address airbome areas or HRAs with dose rates <1000 mrem /hr. Therefore, these areas will be included in the baseline inspection.
Incidents that would be tracked under this Pl include:
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A single nonconformance of TS controls or comparable 10 CFR Part 20 requirements i
applied to all high-radiation areas (HRAs) having dose rates 21000 mrem /hr.
A single nonconformance with 10 CFR Part 20 and/or licensee procedural requirements regarding radiation protection controls associated with VHRAs.
A single occurrence of an uncontrolled exposure in excess of 10 percent (%) of the non-stochastic or 2% of the stochastic dose limits specified in 10 CFR Part 20.
l l 6 election Methodology l
Licensee identified incidents involving the loss of one or more barriers to an HRA, VHRA or airbome area shall be reviewed to determine the potential for an overexposure, if adequate corrective actior;s were taken and that they were not recurrent. Radiologically significant work performed in areas not addressed by the Pl shall be periodically observed to ensure that appropriate physical and administrative controls have been instituted to prevent uncontrolled i
entry. For HRAs < 1000 mrem /hr, the inspection shall verify that the instituted controls are
, comparable to those required by the applicable TS.
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l: l Frequency This inspection shall be performed annually by a Radiation Specialist and as part of the Resident
. Inspector's normal activities.
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INSPECTABLE AREA: Adverse Weather Preparations Scope Inspection activities in this area would focus on the effectiveness of the licensee's program for
- . protecting plant equipment from cold weather and other adverse weather related conditions. The inspection focus would be to ensu*e that risk significant systems and support systems are operated and maintianed within the design assumptions and to ensure that the related
- instrumentation and actuation related components, such as level transmitters, are capable of I
performing their design actuation and control functions, Inspection activities in this area would also include licensee preparations for high temperature, wind, rain, snow, or other site specific
- weather related preparations.
l' Basis
', inspection of this item supports the initiation and Mitigation comerstones by ensuring that 1 l
. systems and components remain capable of performing their design functions. In addition, this I inspection area would have an affect on the Barrier and Security comerstones. !
- The inspection activities are intended to verify that the licensee has taken the necessary steps to demonstrate that the reliability, availability and functional capability of SSCs and associated components are maintained during adverse weather conditions. For example, operating experience indicates that cold weather conditions continue to cause intake structure icing, process line freezing, instrument line freezing, emergency diest.! qenerator oil viscosity problems, essential chiller problems, electrical problems such as grounds, st.curity camera fogging, icing of security perimeter detection devices and icing or rain related problerr.3 with evacuation alarm sirens.
Inadequate freeze protection controls can and have led to inoperable actuation and control systems such as RWST levelinstrumentation and the loss of automatic swithchover, in addition to loss of RWST level indication. Frozen equipment can lead to a common cause/ mode loss of multiple trains and loss of equipment in redundant systems without any indication of a problem until called upon, which would have a significant impact on plant risk. In addition, high temperature conditions can place plant equipment and systems in an unanalyzed condition, which could also have a significant impact on risk.
There are no performance indicators that have been established that can provide results related to the adequacy of the licensee's program for freeze protection or for the adequacy of the licensee's preparations for other adverse weather conditions. Because extemal events are rare, the lack of an initiating event due to an extemal event does not provide assurance that protection against such events is adequate. In addition, due to the rare but possibly risk-important nature of these events, no performance indicators were judged to be suitable to monitor licensee performance in this area.
Selection Methodology Selection of specific inspection activities will be based on the related facility vulnerabilities.
Subsequent inspection activities should be comprehensive enough to ensure that risk significant systems and components, that can be affected by adverse weather conditions, will function as designed during worst case adverse weather conditions. The inspection should ensure that the ,
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licensee has taken the appropriate measures to prepare for adverse weather conditions and during the actual adverse weather periods the inspection should verify that the related adverse weather support equipment is functioning as required.
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The inspection activities should be performed wellin advance to the start of adverse weather conditions,' generally on a yearly basis. During periods with adverse weather, more routine observations of the adverse weather protection equipment, procedures and system testing should be performed.
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lNSPECTABLE AREA ALARA Planning and Controls Scope l This area will verify that the licensee maintains occupational exposure ALARA by properly planning and controlling radiologically significant work activities. Controls, as stated here, refer to those physical (e.g., locked doors, barrier ropes, shielding, engineering controls) and administrative (e.g., surveys, procedures, training, monitoring) barriers that, in the aggregate, serve to mitigate exposure.
Basis l
Inspection in this area supports the Occupational Dose comerstone.
1 This inspection will review whether the licensee meets the requirements of Subpart B to 10 CFR Part 20, which requires that a Radiation Protection program, including procedures and engineering controls, be instituted to maintain occupational dose ALARA.
The NRC position is that the radiological risk (i.e., exposure) to a worker be withir, the l occupational exposure limits defined in 10 CFR Part 20 and ALARA and to minimize the potential for health effects. Effective ALARA planning will ensure that adequate physical and l administrative controls are in place to mitigate exposure during radiologically significant work. ,
Industry experience has identified frequent events where problems in this area have resulted in '
unanticipated exposure or a loss of control of the work activity. Specific attention should be given to Planned Special Exposures and exposures to Declared Pregnant Workers owing to the higher risk involved.
l There is no Performance Indicators established in this area. Assessment of the ALARA program effectiveness is site-specific and highly dependent upon previous operational history, work scope and worker experience.
Selection Methodology Review licensee assessments of the ALARA program to identify recurrent problems or trends that should be reviewed during the inspection. Select jobs having a high individual and/or collective dose or located in an HRA, VHRA or airbome area by attending licensee planning and RP briefings and reviewing RP&C logbook entries and past. outage histories. Compare current -
licensee performance to established exposure goals and previous performance, assess whether
! these goals were aggressive and reasonable, identify what exposure controls were i
implemented, and determine if the licensee's subsequent performance met these goals.
Periodically, obs me selected jobs to determine if the work is being performed as planned.
Frequency
- This inspection should be performed by a Radiation Specialist before and during outages or other periods when a number of radiologically significant activities are scheduled.
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INSPECTABLE AREA: Alert and Notification System Availability Scope:
Inspect the Alert and Notification System (ANS) testing methodology for compliance with NRC guidance.
Basis:
1 This inspection area supports the EP cornerstone and the Facilities and Equipment key attribute.
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The ANS is a risk significant system for notifying the public of the need to take protective actions.
The licensee generally maintains ANS and local govemmental authorities operate it whe'n necessary. Assurance that the system has a high rate of availability increases the assurance
- j. that the licensee can protect public health and safety during an emergency. If an EP program j j consistently ensures that the ANS is in a high state of readiness it indicates that the program is
- operating at or above the threshold of licensee safety performance above which the NRC can .
L allow licensees to address weaknesses with NRC oversight through a risk informed inspection program.
I A Pl, ANS Availability (ANSA) has been developed to measure this performance. However. for the statistics of the Pl to be valid, the testing program must be conducted in accordanca with NRC guidance. The inspection verifies this status.
Selection Methodology The testing methodology at every site would be inspected once during the implementation of the NRC Assessment Program and thereafterif there are changes in the methodology.
l Frequency This inspection would be conducted during the implementation period for the revised NRC assessment program to ensure the standardization of testing and the voracity of the Pl statistics.
The status would be periodically verified during biennial exercises to ensure changes to the testing method continue to meet NRC guidance.
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l INSPECTABLE AREA: . Drill and Exercise Performance (DEP)
Scope:
Collection of data from licensee identified drills, training evolutions, biennial exercise and actual events to support the Pl. Inspect the licensee's self assessment program during the conduct of drills, exercises and other training evolutions and actual events, to verify that the statistics gathered for the DEP Pi represent the actual success rate of performance in risk significant areas.
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i This inspection area supports the EP comerstone and the ERO Performance key attribute.
The implementation of the Emergency Plan is dependant on the performance of the Emergency Response Organization (ERO) in their EP assignments. There are many areas important to Plan implementation, but the most risk significant areas of ERO performance are:
- Timelv and accurate clessification of events
- including the recognition of events as t
potentially exceeding emergency action levels (EALs) and any assessment actions necessary to support the classification.
Timelv and accurate notification of offsite aovernmental authorities; including adequate performance of notifications as specified in the Plan.
Timelv und accurate develooment and communication of orotective action recommendations to offsite authorities: including providing protective action recommendations (PARS) to govemmental authorities, the decision making process to develop the PARS and any accident assessment necessary to support PAR development.
If plant staff assigned EP duties consistently performs these activities in a timely and accurate manner, it indicates that the EP program is operating at or above the threshold of licensee safety performance above which the NRC can allow licensees to address weaknesses with NRC oversight through a risk informed inspection program. The DEP PI has been developed to indicate this performance.
However, for the data provided in the DEP PI to indicate the ERO performance, the licensee self assessment program must determine that the specified activities were performed in a timely and accurate manner. This inspection element would verify the efficacy of licensee self assessment of activities that are included in this statistic.
Selection Methodology Simulated emergency events that are identified in advance of performance as opportunities for the DEP PI may be inspected. This includes the biennial exercise, drills of appropriate scope and operating shift simulator evaluations conducted by the licensee training organization. These inspections could be unannounced. The inspector would observe the performance of the risk L significant areas noted above and determine successes and failures. The licensee assessment
! of the same performance would be observed and compared to that collected by the inspector.
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- Approximately twice per year and during the biennial exercise. The biennial exercise would be observed by regional / headquarters personnel, but the other inspections could be performed by the resident.
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INSPECTABLE AREA: Emergency Action Level Changes Scope:
Inspect changes to the Emergency Action Levels (EALs) since the last inspection.
Basis:
This inspection area supports the EP comerstone and the Procedure Quality key attribute.
t Recognition and subsequent classification of events is a risk significant activity because classification leads to activation of the Emergency Response Organization and notification of govemmental authorities. However, if the EAL scheme is not in compliance with the approved configuration, or changed in accordance with regulations, it will not result in the expected emergency classification.
l l 10 CFR Part 50, Section 50.54 (q) allows changes to the Emergency Plan that do not decrease the effectiveness. However, Appendix E to 10 CFR Part 50, states that NRC will approve EALs.
The inspection review any EAL changes against the requirements of Section 50.54 (q) and
- approve them,if appropriate.
No PI's were established for this area and therefore a baseline inspection is required.
Selection Methodology All changes would be reviewed.
Frequency Performed during the biennial exercise or when EAL changes are submitted to NRC.-
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INSPECTABLE AREA: Emergency Response Organization Augmentation Scope:
Inspect licensee self assessment of Emergency Response Organization (ERO) augmentation.
Generally, this will involve:
review the design of augmentation tests to ensure they provide assurance that emergency response facilities could have been staffed in a timely manner if it had been necessary, a
review the self assessment of test conduct, and review of the results of tests, review the ability of self assessment to identify trends in results, and review the associated corrective actions.
Basis:
1 This inspection area supports the EP comerstone and the ERO Readiness key attribute.
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The licensee system / program to augment the on shift staff with ERO members is a risk i
significant process in EP because the ERO is critical to implementing the Plan in a timely '
manner. This system / process involves a notification system for individual ERO members, training of ERO members in its use and testing to ensure facility activation goals can be met.
i No PI's were established for this area and therefore a baseline inspection is required. l l
Selection Methodology inspection of the test design would be performed for every site initially and periodic inspection of self assessment in this area would be conducted.
Frequency initially for every site during the implementation of the NRC assessment process and periodically there after during the biennial exercise.
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s INSPECTABLE AREA: Emergent Work Scope The inspection activities in this area would focus on the effectiveness of the licensee's controls for the repair of emergent equipment failures that need immediate licensee attention and would not be planned through the licensee's normal 12 week planning schedule. These inspection activities would include a review of related troubleshooting, work planning, establishment of plant
- conditions, tagging, conformance with Technical Specifications and restoration of equipment to service, with an emphasis on verification of plant configuration, both during and following the maintenance activity. The inspection activities would be limited to observing and/or reviewing i
only those potentially higher risk significant emergent activities.
Basis Inspection of this item supports the initiating Events, Mitigation Systems and Security comerstones. In addition, this inspection area would have an affect on the Barrier and Radiological Controls comerstones.
l Inspection activities are intended to verify that the licensee has taken the necessary steps to
- demonstrate that emergent activities are adequately planned, controlled and documented. This 1
would include proper control of troubleshooting and maintenance activities so that appropriate post maintenance testing is identified. This will ensure the continued reliability, availability and functional capability of SSCs and associated components. In addition, emergent failure of equipment can lead to significant risk configurations if other equipment is currently unavailable because of planned maintenance or testing. Proper job planning to minimize risk may involve retuming to service the equipment undergning planned maintenance prior to working on the failed equipment.
J Industry experience has shown that inadequate control of repair activities to equipment during power operation have resulted in plant transients, inoperable safety systems, and/or loss of redundancy. In addition, when the plant is at full power operation, thorough post-maintenance
, testing by the licensee can become more difficult and may warrant additional NRC attention to verify equipment reliability is not jeopardized due to inadequate or inappropriate testing. -
- in addition, the inspection activities should ensure that the licensee has appropriately considered the prioritization and timing of repairs and that the repair activities are factored in with other '
previously planned maintenance or surveillance activities such that overall plant risk is minimized.
3 There are no performance indicators established that provide results related to the adequacy of
- the control or emergent maintenance work activities.
Selection Methodology i
The selection of emergent work activities to be inspected will be based the hierarchy listed in the >
Risk Information Matrices (RIM) for BWR and PWR facilities. The selection from the RIM shall be modified by input from the Senior Reactor Analysis for plant specific insights. This will include selection of activities that have the potential to cause risk significant common mode /cause failures, potential to cause multiple trains or redundant components to be inoperable, potential to i
affect defense in depth, repetitive failures, and/or complex activities involving extensive troubleshooting.
Frequency The frequency of this inspection activity would be based on the number of risk significant problems being experienced by the licensee. Generally, more issues are identified during outage periods.
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INSPECTABLE AREA: Equipment Alignment Scope Inspection activities in this area focus on the effectiveness of the licensee's program for changing the alignment of plant equipment (configuration control) based on procedural and/or operational needs or for clearance control in retuming or removing equipment to/from service for activities such as maintenance, modification or testing. The inspection focus would be to ensure that systems and components are appropriately retumed to service by assessing the effectiveness of the licensee's programs for independent verification, locked valve verification, switching and tagging clearances and system lineups. The inspection activities would be more limited during at power operations with increased emphasis during shutdown evolutions. In addition, much of the inspection focus would be to review the licensee identified issues in this area and to verify that adequate corrective actions are implemented.
Basis inspection of this area supports the Initiating Events, Mitigation Systems and Barrier Integrity Comerstones.
The inspection activities are intended to verify that the licensee has an effective process for maintaining system configuration control, which ensures that the functional capability of the plant system is maintained. Systems and components that are not properly configured may not be capable of performing their intended functions, which results in a loss of availability and funtional capability.
Systems or components that are not properly aligned can lead to the initiation of events, can result in personnel injuries, and can significantly impact the availability and functional capability of plant equipment, which would/could significantly increase the overal1 risk to the plant. It is understood that inspection activities will have minimal impact on reducing the frequency of initiating events. However; a review and documentation of those events does provide valuable assessment information and is an indicator for similar problems within the Mitigation and Barrier comerstones. Inspection activities would normally be expended following emergent work i activities, following risk significant system realignments, or during outage related activities. '
A performance indicator for the unavailability of four systems has been identified. However, !
unavalilability time could/would include unavailability for testing, preventive maintnenance, J predictive maintenance, corrective maintenance in addition to loss of configuration control.
Therefore, the unavailability performance indicator would not be a clear link to the licensee's program or performance related to the control of plant configuration, making it necessary and reasonable to perform inspection activities in this area. In addition, licensees often distribute the configuration problems into multiple site specific performance indicators, such as inadequate j procedure, failure to follow procedure, human performance, or "other", and in some cases the errors in the field are not documented.
Selection Methodology The selection of SSCs to inspect will be based upon the hierarchy listed in the Risk Information Matrices (RIM) for BWR and PWR type facilities. The selection from the RIM shall be modified by input from the Senior Reactor Analyst for plant specific insights. Selection will also be based
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l on the history of previous licensee problems in this area and/or specific scheduled or ongoing risk significant site activities.
l Frequency l Routine inspections would be performed on a quarterly basis if no other inspections in this area i.
have been performed. Inspection activities would normally be expended following emergent work activities, following risk signiiicant system realignments, or during outage related activities.
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9 4 INSPECTABLE AREA: Event Followup Scope Inspections activities in this area would focus on automatic or unanticipated manual reactor trips, significant safety system actuations, and significant plant events, including those in areas such i as radiation control, security or emergency preparedness. Inspection activities in this area v/ould include initial observations of plant and equipment status following trips or actuations and the initial evaluation of significant plant events. In addition, inspection activities would include a detailed review of the licensee's post reactor trip review / evaluation. The intent of the inspection activities would be to ensuro that the risk significance of the event is fully understood, the risk significant issues have bem identified and that the licensee has taken the approrpriate corrective actions.
i Basis This review would verify the actual impact on the Initiation and Mitigation comerstones. In addition, based on a significant issue or event, this inspection area would have an affect on the Security and Radiological Controls comerstones.
Initial review and followup of significant events would be performed to verify that equipment functioned as designed, that personnel actions were appropriate and to determine the risk significance of the event. The inspection activities would verify the availability, reliability and functional capability of risk significant SSCs during actual plant events. The inspection activities
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The inspection activities provide the NRC with a real time assessment of plant and personnel performance for major events and /or issues. This information is needed for the NRC to appropriately address the risk significance of the event and/or issue and to evaluate the need to l expend additional reactive inspeciion resources.
In addition, the detailed review of the licensee's post trip review can be used to assess the licensee's ability to identify and correct any adverse conditions and to ensure that the plant systems and components responded to the transient as designed. Equipment challenged by actual events may lead to operability concems not (scovered by surveillance testing, in the past, this has been found to be very risk significant, for example; with the failure on two occasions of both reactor trip breakers to trip on automatic actuation from the plant's reactor protection system which was not initially identified by the licensee.
A performance indicator for reactor trips is available, however; it would not provide the desired
. real time information. Licensee response to ongoing events provides input regarding the licensee's process, programs and human performance which would not be available from the performance indicator review.
Selection Methodology The inspection activities would include all potentially risk significant " Events" or major " plant issues". Major plant issues would include items such as piping ruptures, fires, significant unmonitored radioactive release, significant security barrier breach and significant personnel
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Frequency.
l The frequency of inspection activities is dependent upon actual licensee performance. In addition, inspection resources can vary greatly based on the issues identified during and following the event and/or significant issue.
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l lNSPECTABLE AREA: Fire Protection Scope l The inspection includes a review of ignition sources, control of combustible materials, and fire l protection systems and equipment. Fire brigade staffing, training and performance as well as l
equipment necessary for plant shutdown following a fire such as emergency lighting, Ap'p endix R diesel generators (when applicable) and remote shutdown equipment would also be included as part of the inspection activities.
- Basis Inspection of this item supports both the initiating events and mitigation systems comer, stones.
The inspection would review licensee controls designed to minimize the probability of a fire and would also review the availability and reliability of equipment necessary to mitigate the effects of a fire.
Proper implementation of the fire protection program is important to provide defense-in-depth against fires by maximizing prevention, detection, suppression, and mitigation capabilities for fires. An effective program reduces the risk of a fire being an initiating event. Also, in the event t
of a fire, reliable detection, suppression and mitigation capabilities ensure the plant can be safely shut down without challenging any of the barriers (fuel, reactor coolant system, or containment).
Plant specific evaluations have shown intemal fires to be high contributors to risk at some plants due to the potential for damaging redundant systems and multiple control circuits and due to the adverse effect on operator mitigation strategies.
There are no performance indicators that assess performance in the area of fire protection.
! Selection Methodology Selection of areas for inspection should be based on the protection of risk-significant areas and systems in the plant as identified in the plant specific fire risk study. For example, inspection for ignition sources and control of combustible materials is particularly important in areas such as the switchgear room, control room and cable spreading room.
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INSPECTABLE AREA: Flood Protection Measures -
t Scope-i Inspection activities in this area focus on licensee's program to protect the plant from potential flooding. These inspection activities would include verification that compensatory measures are documented, equipment is available and staged, and equipment is routinely tested and remains ~ .
fully capable to perform the intended functions. These activities would be performed at those ,
facilities that have the potential for extemal flooding and would also include those facilities with l Intemal flooding concems. ,
i Basis i
This activity would be an input to the initiation and Mitigation comerstones. In addition, this '
i inspection area would have an affect on the Barrier, Security and Radiological Controls comerstones. i Verification of the licensee's implementation of the flood control program would be performed to j insure that the facility is capable of withstanding potential intemal and extemal flooding without undue risk to the safe operation and shutdown of the plant. Flooding would have a significant adverse affect on the functional capability of safety and risk related equipment needed to
- i. maintain the plant in a safe shutdown condition.
Flooding has been shown to be a significant contributor to risk at some facilities. In addition, flooding has the potential to make multiple trains of equipment and support equipment inoperable which would result in a significant increase in risk to the plant. Flooding also has a significant consequence of preventing or limiting operator mitigation and recovery actions.
There are no performance indicators that have been established that can provide results related to the adequacy of the licensee's program for mitigating the consequences for flooding. Due to <
the rare but possibly risk significant nature of flooding events, no performance indicator was I j judged to be suitable for monitoring licensee performance in this area.
- l. Selection Methodology i
' Selection of specific inspection activities will be based on the related facility flooding vulnerabilities. Relative risk information can be obtained through a review of the plant's IPE/ risk study and/or IPEEE. Subsequent inspection activities should be comprehensive enough to ensure that risk significant systems and components will function as designed for the various flooding scenarios.; The inspector should ensure that the licensee has taken the appropriate measures to prepare for both intemal and extemal flooding conditions.
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Frequency The inspection activities at sites having a significant risk potential due to extemal flooding shou!d
- l. be performed wellin advance to the potential start of flood conditions. This would generally be i
performed on a yearly basis for plants with extemal flood potential or for other facilities prior to the start of huricane season.
l The inspection activities at sites having a significant risk potential due to intemal flooding should 1
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i-be performed on a yearly basis. The inspedic.ta:tivities should ensure that operator training is up to date and that flood protection equipenent is installed or staged as necessary, i MCS 11/21/98 i
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l l INSPECTABLE AREA: Fuel F .St Performance l
Scope l l Inspection includes verification of operation of the licensee's capability and performance of in-l plant radio-chemical analyses of the reactor coolant system (RCS).
Basis l
l Inspection of this item supports the cladding performance attribute of the Barrier comerstone.
l l Inspection of fuel cladding performance will provide assurance that the first barrier against l release of radioactivity to the environs is maintained. Failure of fuel cladding would increase the radiation dose to workers and potentially to members of the public.
The fuel cladding integrity is maintained by controlling reactor operation within the established operational limits. Routine sampling and radio-chemical analysis of reactor coolant will detect any fuel cladding failures. Appropriate plant procedures and measures for protecting plant workers from increased dose due to fuel failures and to prevent release of radioactivity to the environs should be implemented.
A performance indicator is provided for RCS activity. This inspectable area could be deleted, if the performance indicator for this area is acceptable and the indicator is verified.
Selection Methodology Frequency i
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l' INSPECTABLE AREA Gaseous and Liquid Effluent Treatment Systems Scope ,
This area will verify that gaseous and liquid radioactive effluent treatment systems (including ,
monitoring instrumentation) are maintained such that radiological releases are properly mitigated, monitored and assessed. The focus is to ensure that abnormal releases are reasonably controlled, that system modifications are properly performed, and that effluent and meteorological monitors are accurate and reliable. Other aspects of system operation (including administrative controls) will be assessed by reviewing licensee assessments and the Annual Environmental Monitoring Report.
Basis inspection in this area supports the Public Dose comerstone.
This inspection will verify that gaseous and liquid effluent processing systems are maintained as .
required by General Design Criteria 60,63 and 64 of Appendix A to 10 CFR Part 50, Radiological l Effluent Technical Specifications (RETS) and the Offsite Dose Calculation Manual (ODCM).
l l The NRC position is that the radiological risk (i.e., exposure) to the public be below the 10 CFR l Part 20 and 40 CFR Part 190 limits and ALARA to minimize the potential for health effr . l Doses below the Appendix I to 10 CFR Part 50 design objectives are considered ALARA by the NRC, Proper operation of the effluent treatment system and monitors will ensure an adequate
- " defense-in-depth" against an unmonitored, unanticipated release of radioactivity to the
!: environment. Overall industry performance has improved, but concems still exist with abnormal releases, system modifications, and monitor operability.
Although a Performance Indicator (PI) was established for this area, it does not address
. abnormal releases, system modifications, and meteorological and effluent monitors. Incidents that will be tracked as a Pl include any effluent release not in accordance with 10 CFR Part 20, Appendix l to 10 CFR Part 50, ODCM and RETS.
Selection Methodology Review licensee assessments, the annual report, and events recorded by the Pl to identify adverse trends, incidents or recurrent problems that should be reviewed. Periodically, walk down the gaseous and liquid treatment systems, radwaste control panels, and effluent and l meteorlogical monitors to confirm that adverse conditions were appropriately addressed and to verify the material condition of equipment. Review the documentation of system modifications and monitor calibrations to verify that they were appropriately performed.
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1 INSPECTABLE AREA: Control Room, Plant and Facility Observations Scope inspection activities in this area would include routine control room and plant walkdowns of all safety significant plant areas, which would include areas with equipment included as part of the maintenance rule, areas included under radiation protection and areas included under security.
This would include routine / periodic walkdowns of areas with safety related equipment; areas containing support equipment for safety related systems; security areas and security boundaries; radiation control areas and radioactive material storage areas; and other areas with equipment used to initiate, mitigate or provide barriers to plant events. Inspection activities would also include a review of control room logs, observations of operator shift tumovers and review or observation of the facility's plan of day meeting and management's review of plant deficiencies.
The inspection activities would focus on identification and understanding of emergent plant issues, potential adverse trends, current equipment problems and /or ongoing activities and their overall impact on plant risk.
The inspector would verify that plant parameters are at expected values; that safety significant systems are appropriately lined up by observing control room or local position indications; that present alarm conditions have been addressed; that ongoing operations, maintenance, and testing activities have been scheduled, approved and are being conducted ir; accordance with l approved procedures; and that system and component deficiencies have been identified and l documented in the licensee's corrective action program.
The primary objective of these inspection activities would be to ensure that the inspectors are aware of current plant and equipment problems and have an appropriate level of understanding of the risk significance of the proposed or ongoing operations, maintenance and testing activities.
These aspects of the inspection effort are important because they will be used as a basis in the risk informed selection process for determining the scope and depth of additional / follow-up inspection efforts. In addition, at each facility appropriate inspection effort will be expended to adequately verify that the licensee is in compliance with the Technical Specification requirements and other license and code requirements.
Basis inspection of this item supports all of the comerstones.
Plant walkdown activities provide the inspectors with up to date plant status and knowledge of scheduled activities. With this knowledge, the inspectors are able to use the information in a risk informed process to plan additional inspection activities and to follow-up on more risk significant issues.
In addition, routine walkdowns provide confidence that systems are being properly maintained and that emergent activities and issues have been adequately addressed by the licensee in such a way that the overall risk impact is minimized. This would also provide confidence that the l scheduled or emergent changes to the plant have not adversely affected the safety functions of the SSCs, which would significantly impact the Initiation, Mitigation and Barrier comerstones.
Emergent plant issues, developing adverse trends and current equipment problems, in addition to scheduled and routine activities, can have an adverse affect on the availability, reliability and
l functional capability of risk-significant SSCs.
This inspection activity will independently verify that the licensee operates the plant safely and
- reliably, that risk significant systems are maintained in an operable condition and that plant activities are performed in conformance with the established requirements.
There are no performance indicators that have been established that can provide results related to safety significance of emerging plant issues or ongoing activities. In addition, performance indicator data would be a lagging indicator and could not include current status of equipment.
Therefore, the baseline inspection program includes inspection of this inspectable area.
Selection Methodology On a more frequent (normal workday) basis, the inspectors will observe and or perform some of
. the following inspection activities: review control room status, review control room logs, and review daily deficiencies reports and /or attend facility plan of day meetings, observe shift l
tumovers and/or attend the plant's deficiency review meetings. The reviews and/or observations will collectively be used to assist in planning additional plant walkdowns or to determine which plant activities to observe and or follow-up.
On a less frequent basis, the inspectors shall walkdown most accessible plant areas. The i.
walkdowns should confirm that adverse conditions are promptly reported, documented and resolved. Walkdowns in areas with equipment or components having a high risk significance should be performed on a more frequent basis than areas with equipment or components with lower risk significance, in addition, areas with ongoing activities that could impact risk should be reviewed more often.
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INSPECTABLE AREA: Heat Exchanger Performance Scope l
The inspection includes a review of the performance monitoring, maintenance and testing activities designed to ensure heat exchanger operability. This inspectable area will be included
- as a sub-part of the engineering plant modification inspection activity. Inspection activities would I
be limited to observations of the licensee's heat exchanger fouling inspections and a review of heat exchanger performance data for the more risk significant systems.
l Basis i
Inspection in this area supports the mitigation systems comerstones by ensuring that the heat exchangers' performance remains adequate to support their design function.
Industry experience has shown that many plants have experienced significant problems with
- degraded performance of heat exchangers' due to problems that include corrosion, silting and
- fouling. Resultant reduction in flow rates and heat transfer capabilities may result in a loss of redundancy of safety systems or a loss of the safety function due to a common mode failure of multiple safety trains. Since the subject heat exchangers never/ rarely see design basis heat
- loads, it is important for the licensee to routinely monitor the performance of the heat exchangers l
to ensure that the heat exchangers are capable of meeting their design requirements.
There are no specific performance indicators established that can provide results related to the
, odequacy of heat exchanger performance.
j Selection Methodology l Selection of equipment forinspection should include those SSCs that are risk-significant and that
) are subjected to the operating conditions and environments most likely to result in degraded performance. The Risk Information Matrix (RIM) and plant specific risk information should be utilized in the selection and prioritization of inspections items. The selection will also be based l on previous related problems, such as reductions in heat transfer capability (fouling), biological growth or clam plugging and/or silting from river water. These items have the potential to cause common cause fai;ures and can have a significant impact on overall plant risk and would indicate a need for additional inspection or review.
l Frequency A review of the heat exchanger performance program would be performed with the annual engineering plant modification's inspection, Based on risk singificance and previous equipment history, on a more routine basis, inspection activities would be provided to observe selected initial heat exchanger opening activities.
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INSPECTABLE AREA: Identification and Resolution of Problems / Issues l
Scope 1
This item will verify that the licensee has an effective problem identification and resolution process. Problem identification and resolution refers to: (1) the deficiency reporting process; (2) licensee self-assessments; and (3) Quality Assurance audits. Additionally, in some plants each department may have its own problem identification and resolution program. The focus of the ,
inspection is on the licensee's effectiveness in identifying, resolving and preventing risk l l significant problems. l Basis Inspection in this area supports each of the comerstones.
The objective of this inspection is to ensure that the licensee effectively assesses performance in the key attributes of each comerstone to correct declining performance prior to having significant events.
An effective problem identification and resolution program is the primary means of reducing risk by correcting deficiencies involving people (i.e., training, knowledge and skills), processes (i.e.,
procedures and programs), and equipment (i.e., design and maintenance) before they manifest in a significant event affecting the health and safety of workers or the public. Industry experience indicates that licensees having an effective program for identifying and resolving problems also have a reduced frequency of events.
There are no Performance Indicators (Pis) that have been established for this item. However, some insight may be obtained from the Pls developed for each comerstone which may reduce the overall inspection effort in this area.
Selection Methodology The inspector shall select a set of outputs from a selected program for review. For cach l comerstone of interest, a sample set comprising licensee assessments and deficiency reports will be selected for review. The selection will be made using information contained in the Risk Information Matrix (RIM) and insights gained from site-specific PRA results, industry experience and NRC inspection findings.
For selected programs, additional issues may be identified by periodic observations of specific activities such as operator simulator training or emergency preparedness, security and fire Protection drills and exercises. These issues shall also be reviewed for inclusion in the sample set.
When reviewing the sample set, consider whether individuals involved in the problem identification and resolution process effectively identify, resolve and correct risk-significant
. problems. Additionally determine if the licensee assessments were adequate in scope to address the key attributes of the comerstone of interest.
Other specific items to consider are:
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If risk insights were used to allocate licensee resources for investigating and correcting i identified deficiencies. :
If root cause analyses and corrective actions were timely and adequate to prevent recurrence.
If the findings were reviewed'as a Pi(if applicable)
If the findings were reviewed for performance trends.
If the findings included industry and NRC generic issues.
Determine if the trend indicated by the sample set selected is consistent with the applicable Pls.
l l Periodically select additional samples of critical elements of the licensee's performance assessment feedback loop if:
Recurrent issues or highly risk significant findings were identified; l .
Adequate corrective actions were not taken in response to a declining trend or performance above a P1 threshold; or The NRC or licensee assessment results indicate risk significant findings that should ;
have been manifested in a negative PI trend. I An observed discrepance between Pl data and NRC or licensee findings is indication that L additional review of PRA assumptions, reverification of applicable Pls and an assessment of changing risk may be required.
! Frequency inspections shall be performed annually by resident and regional specialist inspectors, as l applicable, to assess the overall program (i.e., across all comerstones) and as implemented in each comerstone.
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( INSPECTABLE AREA: Inservice inspection Activities b
Scope i
inspection activities in the area would focus on the effectiveness of the licensee's program for controlling inservice inspection, repair, and replacement of ASME Class 1,2, and 3 pressure retaining components. Inspection activities would include a review of the results of the licensee's steam generator tube inspections, a selected review of risk significant non-code repairs, and at a minimum a review or observation of the reactor vessel ISI examinations at the facility year l
interval, Basis nnspection activities in this area support the Barrier comerstone. In addition, this inspection area l would have an affect on the Initiation and Mitigation systems comerstones.
The inspection activities are intended to ensure that the licensee has an effective program for monitoring degradation of steam generator tubes, controlling non-code repairs to ASME components, and performing the required periodic 10 year ISI examinations, all of which relate to L the continued reliability of the risk significant systems and components.
Degradation of the reactor coolant system, steam generator tubes, or safety related and/or i
' support systems would result in a significant increase in risk. Degraded piping or tubes would jeopardize the integrity of barrier systems and would increase the risk impact due to initiation of events. In addition, it would result in mitigation systems not being capable of performing their intended design functions. Based on these considerations, inspection activities are necessary to ensure that the licensee has an effective ISI program to ensure that risk significant ASME Class 1,2 or 3 piping system degradation is identified and is promptly and appropriately corrected.
There are no performance indicators that have been established that can provide results related to the adequacy of the licensee's program for ensuring system integrity in accordance with ASME requirements Selection Methodology Inspection activities would include reviews of selected steam generator tube non-destructive i examinations.
Inspection activities would include observations and/or reviews of the licensee's reactor vessel y non-destructive examination.
Inspection activities would include risk informed selected reviews of non-code repairs to ASME Class 1,2, or 3 piping systems or components.
Frequency Review or inspection of steam generator tube non-destructive examinations would be performed following each refueling outage.
4-Review or inspection of reactor vessel non-destructive examinations would be performed at the 1
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( Inspection activities related to non-code repairs would be specified when non-code repairs are i performed.
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.4 lo INSPECTABLE AREA: Inservice Testing of Pumps and Valves-ASME Section XI l
SCOPE Inspection activities ir, J.is area would be focussed on the effectiveness of the licer,see's program for testing of pumps and valves as required by ASME Section XI. Inspection activities in l this area would include a review of equipment trend results and observations of selected pump performance testing, valve stroke t!me testing, relief valve setpoint testing, and check valve
! testing. Test failures would be captured for review under the corrective action program inspection activities.
INSPECTION BASIS Inspection activities in this area would provide input to the initiating, Mitigation and Barrier cornerstones.
Inspection of this area would be performed to verify that the required testing is being performed as required and that plant equipment is functioning as designed. Inspection activities in this area are intended to verify the effectiveness of the licensee's program for testing of pumps and valves.Section XI testing program was specifically designed to demonstrate the reliability of l components and to identify degrading components prior to actual failure. The trending of the Section XI test data is necessary to identify degradation of components so that the licensee can initiate prompt corrective actions before the degradation causes a loss of functional capability.
This ensures that equipment will be available and have adequate functional capability if called l~ upon to mitigate the consequences of an accident.
The ASME Section XI pump and valve testing program provides assurance that plant equipment l will function as designed. Degraded equipment, even on less significant systems can collectively have a significant impact on overall plant risk.
There are no performance indicators that have been established that can provide results related to the adequacy of the Section XI testing activities. Therefore, the Baseline inspection program includes inspection of this inspectable item.
SELECTION METHODOLOGY l The selection of SSCs to inspect, related to ASME Section XI pump and valve testing, will be l based upon the hierarchy listed in the Risk Information Matrices (RIM) for BWR and PWR type facilities. The selection from the RIM shall be modified by input from the Senior Reactor Analyst for plant specific insights. Selection will also be based on the history of previous licensee implementation problems in this area, any adverse trends identified within the Section XI pump or valve trending program, and/or following significant maintenance or modification activities on specific components.
FREQUENCY The inspectors should routinely review Section XI pump and valve trending data.
On a monthly basis, observations of pump testing and valve stroke testing evolutions should be performed.
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. valves that cannot be tested during normal operations, should be performed.
Observations of check valve testing and relief valve testing should be performed based on the licensee's schedule for testing.
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! INSPECTABLE AREA: Large Containment isolation Valve Leak Rate and Status l Verification Scope inspection activities in this area would be focussed on the adequacy of the licensee's testing program for large containment isolation valves that provide a direct flow path from the containment atmosphere to outside containment. At most facilities the inspection scope would l
' be limited to the containment purge and ventilation vaives and personnel access hatches.
Inspection activities related to leak rate testing for most of the containment isolation valves and/or containment integrity issues would be captured by the corrective action program inspection activities.'
Basis
' Inspection in this area supports the Barrier comerstone.
The inspection activities are intended to verify that the licensee has an acceptable process for insuring that major containment isolation valves will function as designed in preventing the release of contamination following a design basis accident.
L The normal containment ventilation isolation valves tend to be very large valves with soft rubber seats. Industry experience has shown that the seats tend to dry out over long periods and fail to maintain their leakage integrity. Pressurized water reactor containment purge valves are routinely opened during plant operation to purge the containment or to allow reductions in l containment pressure. The constant cycling results in degradation of the valve seats, in both j cases inspection efforts would be focussed on insuring that the valves continue to meet the l design leakage requirements and that the maintenance and testing efforts are appropriate.
There is a performance indicator for total leakage from all containment penetrations. However, the limited inspection activities detailed in this inspectable area would still be necessary due to i the failure mechanism of the large containment ventilation isolation valves and the failure of the personnel access hatches and containment purge isolation valves caused by routine cycling during power operation.
Selection Methodology l l Selection of inspectable items would be limited to the major containment ventilation and purge l valve and personnel access hatch testing evolutions. Additional review of other containment penetrations would be performed as various problems are identified and es specified in the guidance for the Identification and Resolution inspection area..
Inspection activities of local leak rate testing will be required to verify the containment penetration i leak rate performance indicator. Selection and scope for this inspection activity will be addressed in the Performance Indicator Verification Section.
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Verification of local leak rate testing and the associated inputs to the containment leak rate performance indicator, will be perfomed on a routine basis. This willinclude follow up for leak rate identified valve repairs and selected leak rate testing observations.
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l INSPECTABLE AREA: Licensed Operator Requalification l
1 Scope inspection activities in this area would focus on the effectiveness of.the licensee's program for conducting operator requalification training. Inspection activities would include a review of requalification examinations, administration of requalification examinations, would verify that the facility's operating history has been factored into the requalification program, review the training feedback system, review the remedial training program, and ensure conformance with operator license conditions.
Basis Poor operator performance would/could impact one or more of the initiation, Mitigation, or Barrier
- comerstones. Operator. performance can also be a significant factorin the initiation of events.
Operator performance provides a vital link in mitigating the consequences of improper or i unforseen equipment performance. Degraded operator performance results in increased risk l due to its impact on the human factors terms, assumed operator recovery rates and personnel l induced common cause error rates assumed in the facility IPEs. Human performance errors and failure to recover from accident events are the most risk imponant events at a facility.
There are no performance indicators that have been established that can provide results related to the adequacy of the licensee's licensed operator requalification program.
Selection Methodology Use the basic methodology specified in IP 71001, with revisions specified by regions. . In addition, l resident inspectors should observe and evaluate simulator activities, specifically while the l licensee is simulating risk significant scenarios.
Frequency inspection activities associated with operator requalification would be based on the licnesee's program and testing for licensed operator requalification. Inspection activities would typically be performed yearly.
Routine inspections / observations of simulator activities would be performed on a quarterly basis.
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i INSPECTABLE AREA: Maintenance Rule implementation Scope p The inspection includes a review of goal setting, performance monitoring, repetitive failure E j determinations, and evaluations of functional failures and maintenance preventable functional '
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' failures. The scope of the inspection activities would include those systems covered under the maintenance rule which would also include a review of the licensee's implementation of the l j maintenance rule requirements for those systems.
l Basis inspection of this item supports the initiating event, mitigation systems and barrier integrity I
' comerstones by assessing the effectiveness of the licensee program in ensuring availability and reliability of plant equipment. l l Proper implementation of the maintenance rule is important to ensure reliable operation of plant I equipment within the scope of the rule and where goals have not been met the program has l- been adjusted appropriately. The program should ensure there is a proper balance that optimizes availability and reliability when removing equipment from service for preventive maintenance. High availability and reliability result in a high probability that accident mitigation 1
l systems will perform successfully when needed and barriers will remain effective in preventing !
y the release of radioactivity. !
!- This inspection area supplements the safety system availability performance indicator, in addition, inspection activities in this area would provide an assessment of equipment reliability where a performance indicator does not exist.
Selection Methodology Selection of equipment for inspection should include those SSCs that are within the scope of the maintenance rule. The Risk information Matrix (RIM) and plant specific risk information should I
be utilized in the selection and prioritization of inspections items. Selection of items or issues to review in this area would be based on equipment problems identified in the site's problem identification systems such as problem reports, work requests and/or engineering work requests.
Frequency inspection activities would be based on the frequency of the site's equipment problems that fall within the maintenance rule.
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INSPECTABLE AREA: Maintenance Work Prioritization and Control I
Scope-l Inspection activities in this area would focus on the effectiveness of the licensee's programs for l work prioritization and control during shutdown and operations to reduce risk. Licensee work i prioritization methodologies, work backlog trends, level of maintenance support, and
- assessments of integrated risk of the work backlog would be reviewed by the inspector.
Basis This inspection item primarily supports the Mitigation Comerstone, although both the Initiating; Events and Barriers Comerstones may also be effected.
Maintenance is the primary means of mitigating and managing the effects of component degradation and failures. Operating experience shows that lack of maintenance (component l deficiencies not corrected) or improperly performed maintenance (maintenance activities not well controlled) can greatly contribute to the risk for event initiation, and may cause SSCs to not l- function properly if called upon to mitigate the consequence of an event. Operating experience shows that for risk significant events identified through the Accident Sequence Precursor (ASP) program, work control or failure to maintain equipment represent the majority of causes.
i Appropriate identification, prioritization, planning, scheduling, and completion of work that is the most risk significant is essential to safe operation.
' .There are several established Performance Indicators (Pls) that indirectly infer the quality of work prioritization and control to reduce inspection effort in this area. However, events that have reached the ASP threshold (E-06) tended to be random and were not predicted through existing Pls. -
Selection Methodology Based on PRA insights, the RIM, known licensee performance weaknesses, licensee self-
, assessment results, and corrective action prograrn results, the inspector will assess the
- licensee's program for maintenance work prioritization and control. The inspector should l: understand the licensee's (1) use of PRA and reliability maintenance data, (2) designation of critical SSCs, (3) prioritization schemes for maintenance of critical SSCs, (4) determination of PM and CM frequencies, (5) processes for adjusting risk indicators for maintenance performed during-online and shutdown conditions, (6) integrated risk assessment of taking a risk important SSC out of service for monitoring or maintenance while other supporting or accident mitigation SSCs are not available, and (7) emergent work activities. Emergent work activities may not be well l
planned or appropriately risk-categorized because of their work complexity, the desire to continue plant operations (or retum to plant operation), minimal pre-job briefing, and the lack of ideal plant conditions for rigorous post-maintenance testing. Adjustments to the level of inspection effort E would be made through consideration of work backlog risk, effecfveness of preventive and
- predictive maintenance in preventing failures, corrective maintenance plans for critical SSCs, and the operating condition of the plant, p Frequency I
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INSPECTABLE AREA: Off-Normal Plant Operations-Scope The bspectirn activities will be used to evaluate operator and equipment performance for other I than normal / routine operations. This inspection activity will provide a vital tie between operator performance observed under simulated conditions and those observed during non-routine plant operations. This activity will also provide a snapshot of plant and equipment performance during l
' transient conditions. Inspection activities will be limited to observations of plant shutdowns, selected portions of plant startup activities, and selected downpower evolutions related to equipment testing or repair. In addition to providing observations of non-routine plant operations, observations in this area provide increased opportunities to observe more significant plant l
transients and to evaluate operator and equipment performance dumg those non-simulated transient conditions.
l Basis This inspection primarily supports the mitigating systems comerstone by providing assessment of
- operator performance during transient and off-normal operations. Poor operator performance could also affect the initiating Events and Barrier Integrity comerstones.
Operator performance provides a vital link in mitigating the consequences of improper or unforseen equipment performance. Degrading operator performance results in increased risk due to its impact on human factors terms, assumed operator recovery rates, and personnel induced common cause errors. Probabilistic risk assessments have shown that human errors can be very significant contributors to risk, in particular during recovery from accident events.
Operator performance under abnormal plant operating conditions cannot be sufficiently covered i by a performance indicator.
l Selection Methodology l
'~ All infrequently performed activities that significantly increase the possibility of plant transients or require significant intervention by plant operators, should be considered for inspection. Follow-up inspection of operator and eq tipment performance following a significant plant transient should i
also be performed to assess ue performance of operators and equipment. The items to be l selected should be based on the complexity of the activity and the potential risk significance of possible operator errors and/or equipment problems.
l Frequency All major evolutions should be considered for review and/or observation. Typically a good i
performing plant will only have one startup and one shutoown evolution per year.
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l INSPECTABLE AREA: Operability Evaluations l
Scope l Inspection activites in this area would focus on the effectiveness of the licensee's program for the evaluation of degraded and non-conforming conditions affecting plant systems, structures and components (SSCs). Inspection activities would be limited to a review of those potentially risk significnat degraded and non-conforming conditions affecting SSCs that are considered to be operable and fully capable of perfoming their design functions based on written operability evaluations. Initial reviews of the operability evaluations should be performed following formal i completion of the evalualtions by the licensee.
Basis Inspection of this item primarily supports the mitigating and barrier systems by ensuring risk-significant SSCs are fully functional to perform their design function. In addition, this inspection area would have an affect on the initiating Events, Security and Radiation Control comerstones.
The inspection activities are intended to verify that the licensee has taken the necessary steps to demonstrate that the reliability, availability and functional capability of the SSCs and associated components are maintained although the SSCs ars <k- ..ied and/o non-conforming in some way. ;
As a result of the size and complexity of a nuclear power plant, degraded and non-conforming conditions are frequently identified at all plants. Risk significant SSCs are often affected and the l
degraded or non-conforming condition cannot always be corrected immediately. An improperly evaluated degraded and/or non-conforming condition may result in continued operation with a SSC that is not capable of performing its design function which would result in operation of the plant outside of its design and license bases. The potential effects on safe operation could include the loss of redundancy within a safety system, the loss of safety function or a reduction in the safety margin assumed in the plant design and analyses.
The inspection would ensure that the evaluations include an adequate technical justification to support the operability evaluation and would verify the implementation of any compensatory measures.
There are no performance indicators that provide effective assessment of the quality of operability evaluations.
Selection Methodology l
The selection of operability evaluations forinspection should be based on the information and I guidance provided in the RIM. Also, evaluations for all equipment included in the plant technical .
specifications should be considered for inspection. Screening of all operability evaluations for l l those systems covered by the maintenance rule should be considered. Those operability l evaluations that are potentially risk signifier 71 should be reviewed in detail. !
l Frequency l
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I The frequency of inspection activities is dependent upon actual licensee performance in this area.
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i-INSPECTABLE A:<EA: Operator Work-Arounds l
Scope i
l Inspection activities in this area would focus on plant and control room deficiencies that have the potential to affect operator performance in conductiong routine and non-routine evolutions or the l potential to affect equipment performance during non-routine evolutions or transients. Detailed l inspection activities would be limited to those risk significant deficiencies that could compromise l equipment and personnel mitigation strategies. A more detailed review would be conducted on those items / issues where operator performance during transient or mitigation evolutions would
_ be affected. This would include those deficiencies that would require operator actions for the
! SSCs to meet their design functions.
i i Basis l
Inspection of this area supports the initiating event and mitigating system comerstones. In addition, this inspection area would have an affect on the Barrier, Emergency Preparedness, i Radiological Controls, and Security comerstones.
Operational work-arounds can have an adverse effect on the functional capability of a system to perform its design function in that the system may not be capable of performing its design j ~ function without operator intervention.
- l. Operator work-arounds imply that a system or component is not functioning as designed and that special operator intervention or use of other components would be necessary to complete a required task. Too many operator work-arounds or work-arounds requiring complex operator actions reduce the effectiveness of the operations staff in responding to transient conditions and willincrease the chance of operator errors. PRAs have identified human errors as significant contributors to risk.
' There are no performance indicators that have been established that can provide results related to the adequacy of the licensee's process for controlling operator work arounds. Performance indicators may be able to identify the number of operator work arounds and length of time in place, however; they will not be able to indicate the significance of any one operator work around l item.
l Selection Methodology items should be selected for inspection based on the risk significance of the SSCs using thsoe j systems coverd by the maintenance rule and by using the RIM for guidance and to what extent i the work-around introduces additional burden on the plant operators. The extent that plant procedures may be affected should also be considered when selecting inspection items.
l Frequency The frequency of inspections in this area is dependent upon actual licensee performance in this area.
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, INSPECTABLE AREA: Permanent Plant Modifications I-Scope inspection activites in this area includes the review of design, installation, configuration control,
- and post-modification testing for the potentially risk significant permanent modifications of the l SSCs covered by the maintenance rule, inspection activities would also include an in-depth t
review of changes to the initial licensed design and/or design basis documents.
l Inspection Basis-Inspection of this area supports the design and design control attributes of the Mitigating Systems, Barriers, and initiating Events comerstones. In addition, this inspection area would have an affect on the Radiation Control and Security comerstones.
Inspection of permanent plant modifications provides monitoring of the licensee's performance to
!' ensure that the design bases for risk-significant systems, structures, and components (SSCs) have been maintained and that the changes have not adversely affected the licensing and design
- bases and safety functions of the SSCs. Plant modifications may introduce changes to the assumptions and models used in the plant specific PRA. Modifications to one system may affect the design bases and functioning of other interfacing systeras. Also, similar modifications to several systems could introduce potential for common cause failures that affect plant risk.
industry experience has shown that modifications to risk-significant SSCs can adversely affect their availability, reliability or functional capability. The baseline Inspection of permanent modifications should focus on: (1) compliance with regulations, (2) consistency with defense-in-
' depth philosophy, (3) maintaining sufficient safety margins, and (4) acceptability of the effects on risk.
Verification of post-modification testing to confirm that the objectives of the modification are met l and verification that the system is restored to the required configuration after completion of the
! modification are important. Design requirements that cannot be verified by testing of the modification, such as seismic or environmental qualifications should also be reviewed.
No performance indicators have been established that can provide results related to the adequacy of permanent modifications. Therefore, the baseline Inspection program includes
- l. Inspection of this inspectable area.
Selection Methodology There are sorne SSCs that are not risk-significant; therefore, the baseline inspection program i has a process for selecting those systems that are risk-significant. The baseline inspection should be performed only on those SSCs where there is an increased risk to the safe operation of the plant, where there is potential for compromise of the defense-in-depth philosophy, or l where there is a potential for significant reduction in safety margin. ,
Frequency l
l Inspection activities in this area should be performed on an annual basis. SKM- 11/22/98 1
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INSPECTABLE AREA Physical Protection System (Barriers, intrusion Detection System, and Alarm Assessment) l l'
Scope i
Verify that the licensee has an effective physical protection system in place capable of providing high assurance that the facility is protected against the extemal threat of radiological sabotage. l The system includes protected and vital area barriers, associated intrusion detection systems, and alarm assessment capabilities.
Basis Inspection of this area supports the Physical Protection Comerstone.
l This is a risk significant system that is necessary for protection against the extemal threat of radiological sabotage. Operability of the system is necessary to identify and initiate response to security events. The system is the first line of defense in the " defense in depth" concept of protection against radiological sabotage. The risk significance is based on an exploitable '
vulnerability by a person (s) with the intent and capability of committing radiological sabotage. ,
The frequency of occurrence of this type event has been low. However, the consequences of- '
such an event would be moderate to high.
The performance indicator for this area is that these systems are available and capable of performing their intended function 95% of the time. This indicator is considered adequate to assess performance and no additionalinspection of this area is necessary.
Selection Methodology N/A l Frequency Review the performance indicators quarterfy.
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! INSPECTABLE AREA: Piping System Erosion / Corrosion l
l Scope l l-The inspection activities in this area would focus on the effectiveness of the licensee's erosion and corrosion program. Inspection activities would include reviews of the licensee's monitoring, detection and correction of piping and component degradation caused by erosion and/or corrosion. Inspection activities would ensure that SSCs were being adequately monitored and
! that appropriate corrective actions were implemented. Inspection activities would be limited to the site specific risk significant SSCs and would include reviews of system test results and ,
reviews of corrective actions for identified deficiencies. '
Basis inspection of this item supports the initiating events and mitigation systems comerstones. In addition, this inspection area would have an affect on the Barrier, Emergency Preparedness, j Radiological Controls, and Security comerstones.
l The inspection activities are intended to verify that the licensee has adequately implemented the erosion / corrosion program so that the SSCs remain reliable and fully functional.
Effective implementation of an erosionicorrosion program is important to n,inimize the potential
' for high energy fluid system failures that can result in plant transients, damage to plant-equipment and/or injury of personnel. The industry has experienced failures of steam system
. piping as a result of the effects of erosion / corrosion which underscore the importance of monitoring licensee performanco in this area. Effective implementation of the erosion / corrosion program is also important to minimize the potential for intemal flooding or a loss of system function This includes loss of system or component cooling due to service water system failures.
There are no performance indicators established that can provide results related to the adequacy of the erosion / corrosion program.
l Selection Methodology l Selection of equipment for inspection should include those systems that are most susceptible to erosion / corrosion effects and whose failure would result in the most severe consequences.
Frequency inspection activities in this area would be on an annual basis for reviews of system test results.
Inspection activities for deficiencies identified by the system test results would be based on
! actuallicensee performance in this area.
PJP 11/22/98 I
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INSPECTABLE AREA: Post Maintenance Testing l
l Scope P
l Inspection activities in this area would focus on the effectiveness in the licensee's program for conducting post maintenance testing. The inspection would focus on significant maintenance involving high risk significant systems or components, in areas that have tha potential to cause common mode /cause failures, where repetitive failures indicate programmatic problems, or on :
maintenance activities that have the potential to significantly impact risk. These would also
- incude the areas of emergency preparedness, radiological controls, and security. Inspection ;
activites would focus on verification that the post maintenance testing activites were adequate to '
l verify system operability and functional capability for the maintenance that was performed.
Basis inspection of this item supports the mitigation systems and barrier comerstones. In addition, this ,
inspection area would have an affect on the initiating Events, Emergency Preparedness, '
Radiological Controls and Security comerstones.
This is the only process available to verify that a system or component is reliable and fully functional following maintenance. I l
Post maintenance testing provides the final check that a system and /or component has been retumed to its required design configuration and will perform its design function (s) following completion of maintenance activities. Inadequate maintenance activities that are not detected prior to retuming the equipment to service can result in a significant increase in unidentified risk for the subject system and in common mode /cause failures and potential for loss of function on redundant trains and identical components in other systems.
This inspection activity will include areas not sufficiently covered by performance indicators and will provide performance input into the assessment process.
Selection Methodology The inspectors will use risk informed and performance based judgement to select those post maintenance activities to be observsd or reviewed._ In addition, the inspector will confirm through records review and/or observations that t For selected system, the inspectors will review the maintenance activities in sufficiert detail to determine if the scope of the post maintenance testing activities are adequate to verify the operability of the specific components worked on. he licensee adequately performs the post maintenance activities. The inspector will selectively verify system restoration for critical components in the system. The inspector will ensure that deficiencies are appropriately identified, documented, and resolved.
Frequency The inspector should review at least one post maintenance testing evolution each inspection period. Based on emergent work additional post maintenance testing should be observed l
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! INSPECTABLE AREA: Pre-job Briefings Scope inspections in this area focus on the effectiveness of pre-job briefings. This would include the adequacy of licensee discussions regarding scope of the work activity, expected plant responses, contingency actions, work responsibilities, plant status requirements, and precautions.
' Inspection activities in this area would be limited to observations of risk significant emergent and L or complex evolutions or high risk initiating evolutions Sasis inspection of this item supports the initiation comerstone. In addition, this inspection area would ;
have an affect on the Mitigating Systems and Barrier cornerstones. j
. Inadequate pre-job briefings can result in inappropriate actions, or actions out of sequence, ,
insppropriate responses to changing plant conditions, wrong equipment being taken out of .
service, unintended actuations, plant trips, and unnecessary radiation exposure. Deficiencies in human performence or controlling work processes during maintenance and testing activities have been shown to be a dominant factor in risk-significant events. Poor pre-job briefings tend to cause an increase in risk, and in some instances, could be a significant contributor to risk.
There are no established performance indicators (Pls) that can provide information as to the adequacy of licensee programo for conducting pre-job briefings.
Selection Methodology The selection criteria to determine which work items should be inspected will be primarily based upon the hierarchy listed in the Risk Information Matrices (RIM) for BWR and PWR facilities.
The selection from the RIM shall be modified by input from the Senior Reactor Analyst for plant specific insights. in addition, the selection criteria would include those jobs that: (1) may be technically complicated or involved, (2) involve stressful working conditions such as heat or confined spaces, (3) may involve a high individual or collective dose, and (4) have a short window of opportunity for completion (emergent work). Other considerations for selection would j
' include areas of known poor performance and self-assessment results. Observations of pre-job briefings could include important maintenance, testing, operations, or training activities.
Frequency The frequency or inspection activities is dependent upon actual licensee performance in this area.
RLL 11/22/98 l
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INSPECTABLE AREA Radiation Monitoring Instrumentation
' Scope This area will ensure that criticality, area radiation (ARMS) and continuous air (CAMS) monitors
- .are reliable and accurate during activities that could produce transient, HRAs, VHRAs or airbome l- areas. This inspection will also include the containment dome monitors, because of thei~r
' importance in accident analysis and classification.
l l Basis L
Inspection in this area supports the Occupational Dose comerstone
- l. This inspection will verify that these monitors are calibrated and meintained as required'by 10
- . . CFR part 20 or a licensee's Technical Specifications and procedures.
The NRC position is'that the radiological risk'(i.e., exposure) to a worker be within the occupational exposure limits defined in 10 CFR Part 20 and ALARA and to minimize the potential for health effects. These monitors provide indication of changing radiological conditions to workers such that actions to prevent an overexposere can be taken. Industry experience has identified several events where these monitors were the primary !sdication that radiological conditions had significantly changed as a result of planned or unplanned activities.
There is no Performance Indicator established for this area.
Selection Methodology During the inspection, attend plan of the day meetings, RP briefings and review RP logbook entries to identify those areas where, as a result of ongoing activities, there is a reasonable potential for creating an HRA, VHRA or airbome grea or, in fuel storage areas, of having an inadvertent criticality. Examples of such activities include radwaste resin transfers, movement
- of irradiated material and storage / receipt of new fuel. Additional insights can be obtained from industry events documented in NRC Information Notices or HP Positions or from the generic listing of monitors of interest (summarized by plant type) in Tables 1 and 2 below.
Table 1: Boiling Water Reactor Monitors LOCATION ARMS CAMS TIP Room Monitor x n/a Refueling Floor x x ECCS rooms (primarily RHR) x x Drywell n/a x (undervessel)
Containment x (Dome Monitor) n/a l
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l-Table 2: Pressurized Water Reactor Monitors i
LOCATION ARMS CAMS !
l -- Refueling F!oor x x !
i Containment . x (Domo Monitor) n/a ECCS rooms (primarily RHR) x x If there were indications of fuel leakage, then verify that applicable CAM data was analyzed for alpha emitting radionuclides.
Frequency This inspection shall be performed annually t i a Radiation Specialist and as part of the Resident inspector's normal activities.
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INSPECTABLE AREA Radiation Worker Performance Scope This area consists of observing radiation worker (including Radiation Protection and Chemistry i (RP&C) technicians) performance to verify that they aware of and use appropriate radiological controls when performing radiologically significant work.
Basis inspection in this area supports the Occupational and Public Dose cornerstones.
The objective of this area is to verify that workers understand the radiological hazards associated with nuclear plant operation, effectively identify and control these hazards, identify and resolve adverse trends or deficiencies, and maintain proper oversight of work.
The associated risk is the potential for a significant, unplanned exposure resulting either directly or in part by the failure of a worker to perform a required task owing to poor knowledge or training. Recurrent problems in this area have been identified by the industry as a root or contributing cause in many exposure events and in some events involving the unplanned release of radioactive materialto the environment. This is of special concem during outages, when radiologically significant work is often performed by contract staff having varying levels of expenence.
The established Performance Indicator provides only partialindication of performance for this area and additional inspection is necessary. This indicator is any inadvertent release of radioactive material from the protected area which could reasonably result in public exposure > 1 mrem /yr Total Effective Dose Equivalent.
Selection Methodology During the inspection, observe worker briefings and tumover meetings and review RP&C logbook entries and Corrective Action Program results to identify emergent issues, adverse trends or other problems with worker performance. Additionally, observe and interview workers performing radiologically significant activities to determine their knowledge of radiological hazards (both site specific and industry wide) and required controls. This may include accompanying an RP&C technician on routine rounds or observing a liquid radwaste release. For any identified deficiencies, determine if they were or should have been identified through the licensee's Corrective Action Program and/or tracked under a Pl.
Frequency This inspection shall be performed annually by a Radiation Specialist as part of the ALARA Controls and Planning, Gast ous and Liquid Effluent Treatment Systems and Radioactive Material Processing and Shipment inspections. It shall also be reviewed as part of the Resident inspector's normal activities.
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INSPECTABLE AREA: Radioactive Material Processing and Shipping Scope 1
This item will review the licensee's radioactive waste processing and shipping programs to '
ensure that waste packages are properly classified, placarded / labeled and packaged to ensure the health and safety of workers and the general public.
Basis l
This inspection supports the Public Dose cornerstone.
This inspection will verify that radioactive waste packaged for transport complies with th,e external radiological limits and required placarding, labeling, etc specified in 10 CFR Parts 20 and j 71 and the applicable provisions of DOT regulations 49 CFR Parts 170-189 to limit inadvertent 1 exposure to members of the public. Additionally, radioactive waste intended for burial must be I also comply with the requirements of 10 CFR parts 61.55 - 61.57 for waste classification and stability to reduce the risk of inadvertent radioactive release to the environment and to ensure the )
I health and safety of workers at the waste disposal site.
For Public Dose received from transportation activities, the radiological risk results from direct contact with the radiation emitted from a shipment or from an accident involving a breech of the integrity of the shipping container. The NRC position is that the extemal radiologicallimits and required placarding, labeling, etc specified in 10 CFRs Part 20 and 71, and the applicable provisions of DOT regulations 49 CFR Parts 170-189 constitute an acceptable risk threshold. i For waste intended for burial, additional requirements are stated in 10 CFR 61.55-61.57 to 1 reduce the risk of inadvertent radioactive release to the environment and to protect the health and safety of workers at the waste disposal site.
Radiological risk, as defined by the Intemational Council on Radiation Protection (ICRP) and the NRC is the potential for health effects from exposure to radiation. This risk is defined by the linear-no threshold model which states that the potential for health effects increases linearly with exposure with no minimum threshold. The NRC position is that compliance with the waste classification and shipping requirements in 10 CFR Parts 61 and 71 as meeting the ALARA and regulatory thresholds.
Industry performance in this area has improved with few events involving radioactive waste transport. In particular, the infrequent number of shipping events precluded the development of an effective transportation Performance Indicator. However, there continues to be considerable public and licensee interest in maintaining independent NRC oversight of this area to ensure that adequate protection of the public health and safety is maintained. In particular, industry groups feel that the NRC independent inspection process provides public validation of the integrity of the licensee's Radioactive Waste Processing and Shipping program. Therefore, there is justification in including this inspection in the baseline program.
Selection Methodology The baseline inspection should review licensee assessmnts of the program to verify that the requirements of Subparts B and G to 10 CFR Part 20, Subpart H to 10 CFR Part 71, and the applicable Technical Specifications are being met and that radioactive waste is being properly
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Section X: Program Feedback and Assessment (UNDER DEVELOPMENT]
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classified, surveyed; placarded / labeled and packaged for shipment. Particular emphasis should be given on the 10 CFR Part 61 waste characterization and stability requirements (lli.A.3 and
. Ill.C.5 of Subpart G to 10 CFR Part 20) as industry experience has shown this area has not been well addressed in licensee assessments. The inspection should review the licensee's assessment of minor shipping activities, administrative controls or worker training and qualifications and verify that significant changes to the DOT or NRC transportation requirements were addressed by the licensee. Selected shipping activities having some risk-significance (such as Type B orirradiated fuel shipments) should be observed, including reviewing associated shipping records, to provide independent validation of the shipping program. All transportation events reported to the Commission and/or to the licensee should also be reviewed.
Frequency .
This inspection should be performed once per refueling cycle by a Radiation Specialist.
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L i-INSPECTABLE AREA Radiological Environmental Monitoring Program (REMP)
Scope This area will ensure that the REMP reasonably measures the effects of radioactive releases to the environment and sufficiently validates the integrity of the gaseous and liquid effluent release i
program. The focur is on advarse trends or recurrent problems identified through licensee assessments or the Annual Environmental Monitoring Report and periodic observations of worker and equipment performance. The inspection should not focus on the quality of procedures or
] minor administrative processes as these will be addressed in the licensee assessments.
! Basis inspection in this area supports the Public Dose comerstone.
! This inspection will verify that the REMP is implemented consistent with licensee Technical .
4 Specifications to validate that the effluent release program meets the ALARA principle of Section j IV.B of Appendix i to 10 CFR Part 50.
] The REMP supplements the effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are as predicted by the effluent measurements and modeling of effluent pathways. As such, it serves as the final barrier in assuring that the associated dose from radioactive releases is within regulatory limits. Industry 3 experience has shown that the REMP is often the primary method of assessing the potential risk j from unplanned or unmonitored radioactive releases. Additionally, REMP results have served to allay public concems regarding the actual health effects due radioactive releases associated with power plant operation. Collectively, these coc . ems provide justification for including this area in j- the baseline inspection program.
- There are no Performance Indicators established for this area.
Selection Methodology l
, Review licensee assessments of this area and the annual report to identify adverse trends, !
incidents or recurrent problems that should be reviewed. Specific emphasis should be placed on
' verifying that environmental sampling was representative of the release pathways, that observed
- anomalies in environmental monitoring results were understood and that missed samples and/or d
inoperable sampling / analyses equipment were being properly addressed.
l' Periodically, accompany the REMP technician to observe environmental sample collection and observe the operability of the sampler stations. Additionally, review the calibration and maintenance records for the counting room instruments.
Frequency
- A Radiation Specialist will perform an annual review of licensee assessment and REMP report results and a biennial onsite inspection.
NXS Dated 11/19/98
, 4 INSPECTABLE AREA: Refueling and Outage Related Activities Scope Inspection activities in this area would focus on those outage related activities that have the potential to impact the risk to the plant. These areas include plant cool down, transfer to shutdown cooling, solid operations, drain'down evolutions, fuel handling (core off-load / reload),
mid-loop / reduced inventory operations, containment integrity, plant heat up, reactor startup and physics testing. In addition, the inspection activities would include support systems necessary to mitigate the consequences of shutdown accidents, which includes control of switchyard activities, emergency diesel generator availability and vital power availability. Inspection activities in this area would include activities during forced or planned outages and would not be limited to only refueling outages.
Basis inspection of this item supports the initiating events, mitigating systems and barrier integrity
- comerstones.
The inspection activities are intended to verify that the licensee has taken the necessary steps to minimize potential events, maintain defense in depth, ensure the appropriate SSCs are maintained available to mitigate and contain postulated accidents.
Due to changing plant configuration, combinations of equipment outages can place the plant in a condition where single failures can quickly lead to significant adverse conditions such as core boiling. In addition, operations and maintenance personnel are performing non-routine tasks which have greater risk impact due to the extensive amount of equipment that is usually out of service. These items, along with the fact that the barriers to prevent a radiological release are also degraded, result in a significant increase in risk if not appropriately controlled by the licensee.
There are no performance indicators that have been established that can provide results related to the licensee's performance during refueling outages.
Selection Methodology Aciivities would be selected using the plant specific risk data and the licensee's outage risk evaluations. The licensee's day-to-day shutdown risk assessments should also be used to select activities for inspection because they continuously reassess risk due to changing plant conditions, scheduled activities and out-of-service equipment.
Frequency inspection activities are performed in this area during each refueling outage and selected portions of this inspection activity should be performed as necessary during forced or planned outages.
Ils 11/22/98
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( INSPECTABLE AREA Response to Contingency Events (Protective Strategy and implementation of Protective Strategy)
! Scope l
Verify that the licensee has the capability to protect vital area target sets against the design basis threat. The implementation of the protective strategy includes demonstrating that the strategy works the security force can successfully protect against the design basis threat through drills
! and exercises.
Basis inspection in this area supports the Physical Protection Comerstone.
There is no performance indicator for this inspectable area.
This is a high risk-significant system necessary to protect against the design basis threat of radiological sabotage. The licensee should be able to demonstrate the ability to respond with sufficient force, properly armed, appropriately trained and within the appropriate time frame to protected positions in order to interdict and defeat the design-basis adversary force in order to protect vital equipment necessary for the safe shutuown of the plant.
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The ability of the security force to respond to the design basis threat is contingent upon the intrusion detection system being able to detect; the alarm status being communicated to the alarm stations; the assessment functions (closed-circuit television and lighting) and the training of CAS and SAS operators. .
No' power plants in the United States have been attacked by a terrorist group. However, the consequence to radiologicat sabotage if an attack did occur is high.
Selection Methodology This area will be inspected as part of the baseline inspection program. Validation that the intrusion detection system has no vulnerabilities should be confirmed by observation and testing at specific locations determined by the inspectors. Testing is accomplished by inspectors requesting the licensee conducting the tests.
Inspectors should verify that assessment aids can effectively identify a threat during daylight and darkness, inspectors should verify by observation and testing that all participants responding to contingency events are effectively trained in their responsibilities. Some methods inspectors can use are:
A walkdown inspection of the entire intrusion detection system will be conducted to identify locations most susceptible to covert entry. Perfcrmance testing of the system will be conducted by the licensee and observed by at least two inspectors. One inspector should be located in one of the alarm stations and the other at the alarm zones. Testing should be performance based and challenge the system within its design capabilities.
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I INSPECTABLE AREA: Changes to License Conditions (10 CFR 50.54 and 10 CFR l
50.59)
Scope inspection activities in this area would focus on those non-intent changes to the facility and l licensee programs performed under the requirements of 10 CFR 50.54 and those non-intent i
changes to the facility, procedures, tests or the FSAR performed under the requirements of 10 i: CFR 50.59. The inspection activities would include a review of the licensee's required
[ submittals as specified by 10 CFR 50.54 and 50.59 A more detailed review would be performed on those changes that have the potential to be and/or appear to be intent changes. Examples of inspection areas would include safety evaluations performed by the licensee for permanent and temporary facility modifications, procedure changes, final safety analysis report (FSAR) changes, l and security plan changes.
Basis inspection of this area supports the design and design control attributes of the Mitigating Systems, Barriers, initiating Events and Security comerstones.
Inspection of this item provides monitoring of the effectiveness of the licensee's programs for implementing non-intent changes to facility SSCs, procedues, test programs, FSAR and security ~
plans and ensures that the changes were in accordance with the requirements of 10 CFR 50.54 and 10 CFR 50.59. This would provide assurance tha the changes have not adversely affected the safety functions of the SSCs or reduced the effectiveness of the facility security plans.
No performance indicators have been established that can provide results related to the adequacy of the licensee's program for making non-intent changes to the facility. Therefore, the baseline inspection program includes inspection of this inspectable area.
Selection Methodology 1
inspection activities would include a review of information submitted under the requirements of 10 CFR 50.54 and 10 CFR 50.59. Only those items that are potential intent changes or appear to be intent changes would warrent further review. In addition, in-depth inspection activities should be focussed on only those potentially risk significant changes.
Frequency Initial review activities would be performed at the designated time of the licensee's submittals to the NRC.
Additional detailed review would be dependent upon the licensee's performance in this area.
SKM - 11/22/98 I
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! INSPECTABLE AREA: Safety System Design and Performance Capability
[ Scope Inspection includes review of design bases, final safety' analysis report (FSAR), supporting L calculations, as-built conditions, modi 5 cations, testing, and normal and emergency operations of i
i risk significant systems and interfaces with support systems. This would be an in-depth review of a selected risk significant system or systems.
i l Basis Inspection of this area supports the design attribute of the Mitigating Systems, Barriers, and !
Initiating Events comerstones.
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' ' Inspection of safety system design and performance verifies the initial design and subsequent modifications and provides monitoring of the capability of the selected system to perform its
- design basis functions. The inspection should focus on the design and functional capability of j
i components that are not validated by in-plant testing should be reviewed. Also, seismic and environmental qualifications of the SSCs should be verified. The PRA assumptions and models are based on the ability of the as-built safety system to perform its intended safety function successfully. If the design bases of the system had not been correctly implemented in the j installed system, the operation and test procedures, and the supporting analyses and ;
j calculations, the system cannot be relied upon to meet its design bases and performance l
1 requirements. The design interfaces with support systems, such as cooling systems, ventilation l systems, and instrument air system, should also be reviewed.
I j The baseline inspection should focus on: (1) compliance with regulations, (2) consistency with j
defense-in-depth philosophy, and (3) maintaining sufficient safety margins.
' There are no performance indicators that have been established that can provide results related 1 correct implementation of the design bases in the as-built system and the associated plant documents. Therefore, the baseline Inspection program includes Inspection of this inspectable -
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Selection Methodology The baseline inspection program has a process for selecting those systems that are risk-
, significant. The baseline Inspection should be performed only on those SSCs where there is an 8
increased risk to the safe operation of the plant, where there is potential for compromise of the
, defense-in-depth philosophy, or where there is a potential for significant reduction in safety i margin i Frequency -
Inspection activites in this area would be performed on a bi-annual basis and would
- include / replace the annualinspection acitivities specified under Permanent Plant Modifications.
SKM - 11/22/98 i
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l lNSPECTABLE AREA Security Plans and Procedures (includes Security System i
Modifications)
Scope i
i To assure that changes in the licensee's NRC approved physical security plan (PSP), safeguards l- contingency plan and guard training and qualification plan (T&O) has not decreased the safeguards effectiveness of the plans.
Basis inspection in this area supports the Physical Protection Comerstone.
The licensee may make no changes which would decrease the effectiveness of a security plan, safeguards contingency plan, or guard training and qualification plan without prior approval of the commission. Cnanges that decrease the effectiveness of a plan and are implemented without commission approval can affect the risk significance of the physical protection capabilities of the security organization necessary to protect against the design basis threat of radiological sabotage.
There is no performance indicator for this inspectable area.
Selection Methodology l Frequency i
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INSPECTABLE AREA: Technical Specification Surveillance Testing Scope j
Inspection activities in this area would be focussed on Technical Specification (TS) surveillance <
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' testing and would include reviews and observations of required surveillance testing. The !
inspector would review TS testing results for adequacy in meeting the TS requirements, observe ongoing testing to evaluate human performance, and ensure that appropriate test acceptance criteria is in agreement with facility's design requirements. Surveillance test failures would be captured for review under the corrective action program inspection activities. The inspection activities would verify that the surveillance test actually meets the TS requirements.
Basis inspection of this area ensures that safety systems are capable of performing their safety function and wold support the Mitigation and Barrier cornerstones.
Surveillance activities are required to verify that systems and components are reliable and functionally capable of performing their design function. Technical Specification surveillance testing may or may not in itself be risk or safety significant. However, it is the minimum required testing specified in the facility license and ensures that a conservative safety margin exists for '
system capability. In addition, inspections are necessary to ensure licensee compliance with the NRC regulations and licensing bases. Operating experience has shown that test procedure deficiencies may invalidate previously acceptable test results.
There are no established performance indicators (Pls) that can provide an indication of the adequacy of TS required surveillance testing activities.
Selection Methodology The selection criteria to determine which TS surveillance should be reviewed willinclude information provided in the Risk Information Matrices (RIM) for BWR and PWR facilities, and other site specific information that is available, Numerous NRC Generic Communications have been issued involving inadequate testing activities. Corrective actions taken by licensees to address generic testing deficiencies should be considered during this inspection effort. Since some TS surveillance testing requirements may be less conservative (less safety margin) than others, the inspector should modify the inspection sample from the RIM by input from the Senior Reactor Analyst for plant specific insights. The inspector will determine whether TS surveillance activities are completed as required and that performance of the test meets the requirements specified in the TS. The inspector will also determine whether testing deficiencies are identified, documented and resolved as appropriate. For those instances where the design requirement in the TS is not fully verified through testing activities, the inspector should verify that the TS surveillance testing performed provides assurance that the design requirement can be met.
Frequency On a monthly basis, observations and reviews of TS required surveillance tecting would be performed. RLL 11/22/98 l
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INSPECTABLE AREA: Temporary Plant Modifications Scope inspection activities in this area includes a review of design, installation, configuration control, l
i and post-modification testing for the potentially risk significant temporary modification s of the SSCs covered by the maintenance rule.
l Basis 1
I Inspection of this area supports the design and design control attributes of the Mitigating Systems, Barriers, and Initiating Events comerstones, inspection of temporary plant modifications provides monitoring of the licensee's performance in ensuring that the design bases for risk-significant systems, structures, and components (SSCs) have been maintained and that the changes have not adversely affected the safety functions of the SSCs, Temporary modifications may introduce change to the assumptions and models used in the plant specific PRA. A temporary change to one system may affect the design bases and safety functions of other interfacing safety systems. An increase in the likelihood of the occurrence of an initiating event could result from a temporary change. Also, similar temporary modifications to several systems could introduce potential for common cause failures that affect plant nsk.
Industry experience has shown that temporary modifications to risk-significant SSCs can adversely affect their availability, reliability or functional capability. Verification that all safety functions of the system are restored after completion of the temporary modification is important.
No performance indicators have been established that can provide results related to the adequacy of temporary modifications. Therefore, the baseline Inspection program includes inspection of this inspectable area.
Selection Methodology There are some SSCs that are not risk-significant; therefore, the baseline Inspection program has a process for selecting those systems that are risk-significant. The baseline Inspection should be performed only on those SSCs where there is an increased risk to the safe operation of the plant, where there is potential for compromise of the defense-in-depth philosophy, or where there is a potential for significant reduction in safety margin.
Frequency The frequency of inspection activities is dependent upon actual licensee performance in this area.
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C. Process for selecting activities to be inspected within inspectable areas.
The process is depicted in Figure 1, attached. Each year, the baseline inspection program will be developed by Regional managers for each power reactor site.
The inpsectable areas and number of activities, frequency of act!vities and how much time to expend inspecting each activity are fixed in the program using the RIM. The Regional managers may select the type activities to inspect using a risk informed table, the Risk Hierarchy Chart, and site specific information.
ElapIn_1 CONCEPT - WHAT AND HOW TO INSPECT PROCESS TOOLS STEP 1: BASELINE PROGRAM SPECIFIES FOLLOWING FOR EACH INSPECTABLE AREA:
RISK
- FREQUENCY c INFORMATION
- HOW MANY ACTIVITIES MATRIX
- HOW MUCH TIME STEP 2: SELECT RISK SIGNIFICANT GENERAL ACTIVITIES TO MATCH LEVEL RISK OF EFFORT DEFINED IN HIERARCHY STEP 1 CHART (IST CUT) o SITE SPECIFIC INFORMATION (MODIFIED CUT)
STEP 3: CONDUCT INSPECTION o SET OF USING PROCEDURE PROCEDURES
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l D. Process for determining frequency, how many activities and how much time to l l expend inspecting within each inspectable Area l This process was made using information placed in a Risk Information Matrix (RIM). the information was collected from risk analyses in the IPE and IPEEE documents for each reactor site, from defense-in-depth logic, and from history of problems / experience. This RIM is contained in Section VI of this program document.
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l Section Ill: PART 2 of the Program: Process for verification of performance indicators l r
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w e Section IV: PART 3 of the Program: Process for evaluating problem identification and resolution programs, in general, licensees identify problems (conditions adverse to quality) by three processes:
- 1) problem reports or condition reports that are initiated by plant personnel when they observe problems; 2) licensee self-assessments of individual departments (such'as engineering, operations, and radiation control); and, 3) quality assurance audits.
l Problems identified by any of these processes are assessed by the licensee, root causes l are determined, and corrective actions are implemented under a plant-wide corrective action program. At some plants, each department may have its own problem identification and corrective action program.
The NRC will review the licensee's activities in this area to verify that: 1) the lice'nsee's identification and resolution program scope bounds the key attributes in the comerstone; ;
- 2) root causes of problems and issues have been properly determined and corrective actions are timely and effective; and, 3) generic implications or extent of condition have l
been considered. If the NRC review indicates that for any of the key attributes the t
licensee has not been identifying and correcting problems, additional inspections in that are may be proposed.
The NRC program to review activities in this are is composed of two parts. Part one is an annual review of the overall problem identification and resolution programs across all comerstones. Part two is an inspectable area applied to each comerstone.
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I Section V: Inspection Procedures l [UNDER DEVELOPMENT]
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O e Section VI: Interface with other NRC processes (UNDER DEVELOPMENT)
v c Section Vil: Risk Information Matrices (RIM)
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Section Vill: Definitions of terms used in the baseline inspection program (UNDER DEVELOPMENT) j l
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1 Section IX: Projected resources for baseline inspection program (UNDER DEVELOPMENT) l l
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