ML093100167

Re in the Matter of David Geisen, Exhibits - Volume 1, Exhibits 1 - 70, Staff Exhibits 1 Through 20

ML093100167
Person / Time
Site:	Davis Besse
Issue date:	09/09/2009
From:	NRC/SECY
To:	Atomic Safety and Licensing Board Panel
SECY RAS
References
ASLBP 06-845-01-EA, IA-05-052
Download: ML093100167 (390)

Text

UNITED STATES AMERICA STATES OF AMERICA NUCLEAR REGULATORY NUCLEAR REGULATORY COMMISSION COMMISSION BEFORE THE ATOMIC ATOMIC SAFETY AND LICENSING BOARD Matter of In the Matter )

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DAVID GEISEN DAVIO ) IA-05-052 Docket No. IA-'OS-OS2

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)) ASLBP No. 06-84S-01-EA 06-845-01-EA Exhibits --Volu

• Exh~ibits Volume me. 1 Exhibits 1 -- 70 70 Office of the Secretary

J This electronic This electronic text represents represents the Commission's current the Commission's Enforcement Policy. In NRC Enforcement current NRC In a notice notice published in the Federal published in the Federal Register on March 16, 2005, March 16,2005, the Commission announced Commission announced its its intent to use NRC public use the NRC public WebWeb site site and and the Document Access Agencywide Document the NRC's Agencywide Access and and Management Management SySlftll$~5itJrqrg~unicate Sys .cate its "General Statement of Policy and Procedure for NRC its "General Statement of Policy and Procedure NRC En;(<ffi:!~r§~I~~lG~

E n Enforcement Policy,"

Enforcement Policy," to discontinue paper document, publication of the paper discontinue publication document, N~~P6ffir,~rr~l*,Pnnplify a n*, o s~implify the official statement title. The NRC is policy statement official policy is taking these atWl~M~{6~~~

• L4 ,ll t electronically on the NRC available electronically statement is available j statement NRC public Web Web site site and

~~R as the "NRC Enforcement is widely *UJR Enforcement Policy."

Policy."

NRC ENFORCEMENT ENFORCEMENT POLICY Table of Contents Contents Preface Preface I.I. INTRODUCTION AND PURPOSE INTRODUCTION PURPOSE II. STATUTORY AUTHORITY STATUTORY AUTHORITY AND PROCEDURAL PROCEDURAL FRAMEWORKFRAMEWORK A. Statutory Statutory Authority R Prne-PtdirA1 ]rnimouxxtrl-B. Procedural Framework u.s.

U.S. NRC G In re DAVID

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f' r GEISEN )iOv-t-r . . . . e._ _

. a. I I In re DAVIDGEISEN III. RESPONSIBILITIES RESPONSIBILITIES Docket # 1A-05-052 . j/l

1. 1A-05-052 )6 '82 5 IV.

TV. SIGNIFICANCE OF VIOLATIONS SIGNIFICANCE VIOLATIONS ate Date D Marked Marked for ID~

IDJ-.--. 2008 2008 (Tr. p...

p. -" . )

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A. Assessing Significance Significance Date Offered in Ev: I Date Offered #. , 2008 (Tr. p..p .

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2.

Actual Actual Safety Consequence Potential Safety Consequence Potential

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Through WitnessJPnel: anel: lAl/v; ~

Consequence ~

Aci= REJECTED WITHDRAWN REJECTED WITHDRAWN 3.

4. Willfulness Regulatory Process Impacting the Regulatory ome; l1U-.

Daw. i201 *2008 (Tr. p.~

2 O(Tr- P.~~ ~U )

5. Sif,'11ificance Determination Process Significance Determination B. Assigning Severity Level V. PREDECISIONAL ENFORCEMENT PREDECISIONAL ENFORCEMENT CONFERENCES CONFERENCES VI. DISPOSITION OF VIOLATIONS DISPOSITION VIOLATIONS A. Non-Cited Violation 1.1. Power Reactor Licensees

2. - 7. [Reserved]

2. DOCKETED DOCKETED USNRC USNRC 8.

8. All Other Licensees B. Violation Notice of Violation September 9, September 9, 2009 2009 (11(11:00am)

OOam)

C. Civil Penalty OFFICE OF OF SECRETARY SECRETARY OFFICE 1.1.. Base Civil Penalty RULEMAKINGS AND RULEMAKINGS AND ADJUDICATIONS STAFF STAFF 2.

2. Civil Penalty Assessment ADJUDICATIONS CkLM4 ý -,ý) ý P(4

represents the electronic text represents This electronic the Commission's current NRC Commission's current Enforcement Policy. In NRC Enforcement In a notice notice Federal Register on March published in the Federal published March 16,16, 2005, announced its Commission announced 2005, the Commission intent to its intent use use the public Web the NRC public Web site and Agencywide Document NRC's Agencywide and the NRC's Access and Management Document Access Management Syst:ml$~6it:1/tlrQ~unicate its "General SyjmA*-*,rnmunicate Statement of Policy "General Statement Procedure for NRC Policy and Procedure NRC ErR~~r§~I~~1Ba En f*eq1ii'L Enforcement Policy," to discontinue Enforcement Policy," publication of the discontinue publication paper document, the paper document, N~~P6rn:r,~5Jl~~.:IPnnplify official policy Oan93ako sOinplify the official statement title. The NRC is taking these policy statement a~Ot!s.

1gOIt I~/;6~~~

§L~~t J* statement is available statement electronically on the NRC public Web site and available electronically widely ~~R is widely , as the "NRC Enforcement Policy."

NRC Enforcement Policy."

ENFORCEMENT POLICY NRC ENFORCEMENT Table of Contents Preface Preface I.

1. INTRODUCTION AND PURPOSE INTRODUCTION PURPOSE II. STATUTORY AUTHORITY STATUTORY AUTHORITY AND PROCEDURALPROCEDURAL FRAMEWORKFRAMEWORK A. Statutory Authority Statutory Rt Prnrpii~rnl Framewonrl B. Procedural Framework U.S. NRIC In NRO In re DAVID GEISEN) -t-DAVIDGEISEN

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Ill.

III. RESPONSIBILITIES RESPONSIBILITIES Docket # 1A-05-052 1A-05-052 :/J J. 2\7 S ate Marked Date Marked D for ID~

IDi.-,

6 2008 (Tr.P-LL (Tr. p. & )}

IV.

TV. SIGNIFICANCE OF VIOLATIONS SIGNIFICANCE VIOLATIONS f ~U

==.

1'1)

A. Assessing Significance Date Offered in Ev: ~,

Date Offered . 2008 p. L %L))

(Tr. p.

2008 j"r.

1I.. Actual Safety Consequence

2. Consequence Potential Safety Consequence ThroughWitn:

Through s anel: e REETE tV / & WI AcUon:Actfl

~MJT~REJECTED REJECTED WITHDRAWN~

WITHDRAWN 3.

4.

Impacting Willfulness Willfulness Regulatory Process Impacting the Regulatory Date: llU-.Lo . 2008 oe(Tr.(Tr. pPi ~ Lo )

5.

5. Determination Process Significance Determination B. Assigning Severity Level V. PREDECISIONAL ENFORCEMENT PREDECISIONAL ENFORCEMENT CONFERENCES VI. DISPOSITION OF VIOLATIONS DISPOSITION VIOLATIONS A. Non-Cited Violation I.

1. Reactor Licensees Power Reactor

2. - 7. [Reserved] DOCKETED DOCKETED USNRC USNRC 8.

8. AD Other Licensees .

All B. Notice of Violation September 9, September 9, 2009 (11:00am) 2009 (11 :OOam)

C. Civil Penalty OFFICE OF SECRETARY OFFICE OF SECRETARY 1.

1. Base Civil Penalty RULEMAKINGS AND RULEMAKINGS AND ADJUDICATIONS STAFF STAFF 2.

2. Civil Penalty Assessment Assessment ADJUDICATIONS

a.a. Initial Escalated Action Initial Escalated Action b.b. Credit for Actions Related Credit for Actions to Identification Related to Identification c.

c. Credit for Credit Prompt and for Prompt and Comprehensive CorrectiveAction ComprehensiveCorrective Action

• d.

'd. Exerciseq,of Discretion Exercis~of.oiscretion-:

Orders:.

D.

D. _Orders ____ "_;"" :,,";_"0",: -,:__ ,

E.E. RelatedAdministrative Related Actions` - - -

AdriiinistrativeActioris-VII.

VII. EXERCISE OF EXERCISE OF DISCRETION DISCRETION A.

A. Escalation of Escalation Enforcement Sanctions ofEnforcement Sanctions 1.

1. Civil Penalties Civil Penalties 2.2. Orders Orders 3.

3. Civil Penalties Daily Civil Daily Penalties BB. Mitigation ofEJ1.iorcement Sanctionsc Mitigatinn of Enfnrcement Sanctions 1.

1. [Reserved]

[Reserved]

2.2. Violations Identified During Violations Identified Shutdowns or Extended Shutdowns During Extended Work Stoppages or Work Stoppages 3.

3. Violations Involving Old Violations Involving Design Issues Old Design Issues 4.

4. Identified Due Violations Identified Violations Due to to Previous Enforcement Action Previous Enforcement Action 5.

5. Violations Involving Certain Violations Involving Discrimination Issues Certain Discrimination Issues 6.

6. Violations Involving Special Violations Involving Special Circumstances Circumstances' C.

C. Notice of Notice ofEnforcement Discretion for Enforcement Discretion for Power Reactors and Power Reactors and Gaseous Diffusion Plants Gaseous Diffusion Plants

. III. ENFORCEMENT

,,------ c"VIII.' ENFORCEMENT ACTIONS ACTIONS INVOLVINGINVOLVING INDIVIDUALS INDIVIDUALS IX.

IX. INACCURATE AND INACCURATE INCOMPLETE INFORMATION AND INCOMPLETE INFORMATION X.

X. ENFORCEMENT ACTION ENFORCEMENT ACTION AGAINST.AGAINSTNON-LICENSEES NON-LICENSEES XI.

XI. REFERRALS TO REFERRALS THE.DEP ARTMENT OF TOTHE.DEPARTMENT OF JUSTICE JUSTICE

, ':::'-..~~"': :~: '~.'. It*

• XII.

XII. PUBLIC DISCLOSURE PUBLIC DISCLOS\JiutOF ENFORCEMENT ACTIONS OF ENFORCEMENT ACTIONS XIII. REOPENING CLOSED REOPENING ENFORCEMENT ACTIONS CLOSED ENFORCEMENT ACTIONS SUPPLEMENTS -- VIOLATION SUPPLEMENTS VIOLATION EXAMPLES EXAMPLES INTERIM ENFORCEMENT INTERIM ENFORCEMENT POLICIES POLICIES Interim Enforcement Policy for Generally Interim Enforcement Licensed Devices Generally Licensed Containing Byproduct Devices Containing Byproduct Material (10 Material CFR 31.5)

(10 CFR Enforcement Policy Interim Enforcement Interim Policy Regarding Discretion for Enforcement Discretion Regarding Enforcement Certain Fitness-for-for Certain Fitness-for-Duty Duty Issues (10 CFR Part (10 CFR Part 26) 26) 2

Interim Enforcement Policy Interim Enforcement Policy Regarding Enforcement Enforcement Discretion Discretion for Certain Fire Protection Protection Issues (10 (10 CFR 50.48) 50.48)

Interim Enforcement Policy Interim Enforcement Policy Regarding the Use of Alternative Dispute Resolution Resolution PREFACE PREFACE The following policy statement statement describes the enforcement enforcement policy and procedures procedures that the U.S. Nuclear Regulatory Commission (NRC or Commission) and its staff intends to follow in initiating enforcement actions in response to violations of NRC requirements.

initiating and reviewing enforcement This statement statement of general general policy and procedure is publically available available on the NRC public Web site and the NRC's Agencywide Document Access and Management Management System (ADAMS)

(ADAMS) to foster its widespread widespread dissemination. However, this is a policy statement statement and not a regulation. The Commission may deviate from this statementstatement of policy as appropriate appropriate under the circumstances circumstances of of a particular particular case.

I. INTRODUCTION INTRODUCTION AND AND PURPOSE PURPOSE The Atomic Energy Energy Act of 1954,1954, as amended, establishes establishes "adequate protection" as the standard standard of safety on which NRC regulations regulations are based. In the context of NRC regulations, safety safety means avoiding undue risk or, stated another way, providing reasonable assurance of assurance of adequate adequate protection of workers and the public in connection connection with the use of source, byproduct source, byproduct and special nuclear nuclear materials.

While safety is the fundamental regulatory regulatory objective, compliance with NRC requirements objective, compliance requirements plays an important role in giving the NRC confidence confidence that safety is being maintained. NRC requirements, requirements, including including technical specifications, specifications, other license conditions, orders, and regulations, have been designed to ensure adequate adequate protection corresponds to "no undue risk to protection -- which corresponds public health and safety" -- through acceptable acceptable design, construction, operation, maintenance, maintenance, modification, and quality assurance measures. In the context context of risk-informed regulation, compliance compliance plays a very important role in ensuringensuring that key assumptions used in underlying risk risk and engineering analyses remain valid.

While While adequate protection protection is presumptively presumptively assured by compliance compliance with NRC requirements, circumstances may arise where new information requirements, circumstances information reveals that an unforeseen unforeseen hazard exists or that there is a substantially greater greater potential for a known hazard to occur. In such situations, the NRC has the statutory statutory authority to require licensee action above and beyond existing regulations regulations to maintain the level of protection necessary to avoid undue risk to public protection necessary health health and safety.

The NRC also has the authority to exercise discretion operations --

discretion to permit continued operations despite the existence existence of a noncompliance noncompliance -- where the noncompliance noncompliance is not significant from a risk perspective perspective and does not, in the particular circumstances, pose an undue risk to public health particular circumstances, 3

and and safety. When When noncompliance noncompliance occurs, occurs, the NRCNRC mustmust evaluate evaluate the degree degree of risk risk posed posed by by that that noncompliance to determine if specific immediate action noncompliance determine specific immediate action required. Where needed to is required. Where needed to ensure adequate adequate protection protection of of public public health health and safety, the the NRC NRC maymay demand demand immediate immediate licensee licensee action, including a shutdown up to and including shutdown or cessation cessation of of licensed licensed activities.

activities.

Based Based on the NRC's evaluation evaluation of noncompliance, noncompliance, the the appropriate appropriate action action could could include refraining refraining from takingtaking any action, takingtaking specific specific enforcement enforcement action, issuing orders, orders, or providing input to other input other regulatory regulatory actions or or assessments, such such as increased increased oversight oversight (e.g.,

(e.g., increased increased inspection). Since inspection). Since some requirements requirements are more more important important to safety safety than than others, others, the NRC endeavors endeavors to use aa risk-informed risk-informed approach approach when when applying applying NRCNRC resources resources to the the oversight oversight of of licensed licensed activities, activities, including including enforcement enforcement activities.

activities.

The primary primary purpose purpose of the NRC's EnforcementEnforcement Policy is to supportsupport the NRC's overall overall safety mission in protecting safety protecting the public health and safety safety and the environment. Consistent with that purpose, the policy endeavors to:

• 0 Deter noncompliance noncompliance by emphasizing emphasizing the importance importance of compliance compliance with NRC requirements, and

• 0 Encourage Encourage prompt comprehensive correction prompt identification and prompt, comprehensive correction of violations of of NRC requirements.

requirements.

Therefore, licensees,'

Therefore, contractors,2 and their employees who do not achieve the high licensees,) contractors,2 standard of compliance which standard which the NRC expects will be subject to enforcement enforcement sanctions. Each enforcement action is dependent enforcement dependent on the circumstances circumstances of the case. However, in no case will licensees licensees who cannot achieve and maintain adequate adequate levels of safety be permitted permitted to continue continue to to conduct licensed activities.

conduct

'This policy primarily addresses the activities of NRC licensees and applicants for NRC licenses. However, this lThis policy provides for taking enforcement enforcement action against non-licensees non-licensees and individuals in certain certain cases. These non-licensees include contractors and subcontractors, holders of, of, or applicants for, NRC approvals, e.g., certificates certificates of of compliance, early site permits, or standard design design certificates, certificates, and the employees of these non-licensees. Specific enforcement action against guidance regarding enforcement against individuals individuals and non-licensees is addressed addressed in Sections VIII and X, respectively.

2The term term "contractor" 2The "contractor" asas used used in in this this policy policy includes includes vendors vendors who who supply supply products products or services to be used in an an NRC-licensed facility or activity.

4

II. STATUTORY

11. STATUTORY AUTHORITY AUTHORITY AND PROCEDURAL FRAMEWORK AND PROCEDURAL A. Statutory A. Statutory Authority Authority The NRC's enforcement enforcement jurisdiction is drawn from the Atomic Energy Act of 1954, as amended, and the Energy Reorganization Reorganization Act (ERA) of 1974, as amended.

Section Section 161 of the Atomic Energy Energy Act authorizes the NRC to conduct inspections and investigations investigations and to issue orders as may be necessarynecessary or desirable to promote the commoncommon defense and security security or to protect health or to minimize minimize danger to life or property. Section 186 186 .

authorizes authorizes the NRC to revoke licenses under certain circumstances (e.g.,

certain circumstances (e.g., for material false statements, in response statements, response to conditions that would have warranted warranted refusal of a license on an an original application, for a licensee's licensee's failure to build or operate operate a facility in accordance accordance with the terms of the permit or license, and for violation of an NRC regulation). Section 234 authorizes the NRC to impose civil penalties not to exceed $100,000 $100,000 per violation per day for the violation of certain specified licensing provisions provisions of the Act, rules, orders, and license terms implementing implementing these provisions, and for violations for which licenses can be revoked. In addition to the enumerated enumerated provisions in section 234, sections 84 and 147 authorize authorize the imposition of civil penalties for violations of regulations implementing implementing those provisions. Section Section 232 authorizes authorizes the NRC to seek injunctive injunctive or other equitable relieffor relief for violation of regulatory requirements.

Section 206 of the Energy Reorganization ofthe Reorganization Act authorizes authorizes the NRC to impose civil penalties penalties for knowing and conscious failures to provide provide certain safety information information to the NRC.

Notwithstanding Notwithstanding the $100,000

$100,000 limit stated in the Atomic Energy Energy Act, the Commission Commission may impose higher civil penalties penalties as provided provided by the Debt Collection Improvement Improvement Act of 1996.

Under the Act, the Commission Commission is required to modify civil monetary penalties penalties to reflect inflation.

The adjusted maximum maximum civil penalty amount is reflected in 10 CFR 2.205 and this Policy Policy Statement.

Chapter Chapter 18 of the Atomic Energy Act provides for varying levels of criminal penalties (i.e.,

(i.e., monetary fines and imprisonment) imprisonment) for willful violations of the Act and regulations or orders issued under sections sections 65, 161(b), 161(i),

65, 161(b), 161(i), or 161(0) 161(o) of the Act. Section 223 provides provides that criminal penalties penalties may be imposed on certain certain individuals employed by firms constructing or supplying supplying basic components components of any utilization utilization facility if the individual knowingly and willfully vio violates lates NRC requirements requirements such that a basic component component could be significantly significantly impaired. Section 235 235 provides provides that criminal penalties may be imposed on persons who interfere interfere with inspectors.

inspectors.

Section 236 provides that criminal penalties may be imposed on persons who attempt attempt to or cause cause sabotage sabotage at a nuclear facility or to nuclear fuel. Alleged Alleged or suspected suspected criminal violations of the Atomic Energy Act are referred referred to the Department Department of Justice for appropriate appropriate action.

5

B. Procedural Procedural Framework Framework Subpart B Subpart B of of 10 CFR Part Part 2 of of NRC's regulations regulations sets sets forth the the procedures procedures thethe NRC uses in exercising enforcement authority. 10 CFR 2.201 exercising its enforcement 2.201 sets forth forth the procedures procedures for for issuing issuing Notices of Violation.

Notices The procedure procedure to be used in assessing assessing civil penalties penalties is set forth in 10 CFR 2.205.2.205. ThisThis regulation provides provides that the civil penalty process civil penalty process is initiated initiated by issuing a Notice of Violation Violation and Proposed Imposition Proposed Imposition of of a Civil Penalty. The licensee or other other person person is provided an opportunity opportunity contest the to contest the proposed proposed imposition of a civil penalty penalty in writing. After After evaluation evaluation of the the response, response, the civil penalty penalty may be be mitigated, remitted, remitted, or imposed. opportunity is imposed. An opportunity is provided provided for a hearing if a civil penalty is imposed. If a civil civil penalty following aa hearing or if a penalty is not paid following hearing is not requested, the matter matter may be referred referred to the U.S. Department of Justice to institute U.S. Department institute a civil civil action inin District Court. .

procedure for issuing an order to institute a proceeding The procedure proceeding to modify, suspend, or revoke a license license or to take other action against against a licensee licensee or other person subject subject to the jurisdiction jurisdiction of the Commission is set forth in 10 CFR 2.202. The licensee or any other person adversely affected by adversely affected the order may request a hearing. The NRC authorized to make orders immediately NRC is authorized effective if immediately effective if required to protect public health, safety, or interest, protect the public Section 2.204 interest, or if the violation is willful. Section sets out the procedures procedures for issuing aa Demand for Information Information (Demand)

(Demand) to a licensee licensee or other other person subject to the Commission's jurisdiction jurisdiction for the purpose purpose of determining whether an order order enforcement or other enforcement action should be issued. The Demand does not provide provide hearing rights, as only information information is being sought. A licensee licensee must answer a Demand. An unlicensed person may answer a Demand by either providing information or explaining why the Demand either providing the requested information should not have been issued.

III.

I11. RESPONSIBILITIES Operations (EDO) and the principal Executive Director for Operations The Executive principal enforcement officers of of the NRC, the Deputy Executive Executive Director for Reactor Programs (DEDR)and the Deputy Executive Director for Materials, Research Research and State Programs (DEDMRS) have been delegated delegated 3

3 the authority to approve or issue all escalated enforcement enforcement actions. The DEDR is responsible to enforcement programs. The Office of Enforcement the EDO for NRC enforcement Enforcement (OE) (OE) exercises oversight of and implements the NRC enforcement enforcement program. The Director, OE, acts for the Deputy Executive Director in enforcement enforcement matters in his absence absence or as delegated.

3 The term "escalated 3The enforcement action" as used in this policy means a Notice of Violation or civil penalty for "escalated enforcement any Severity Level I,I, II, (or problem); a Notice of 1I, or III violation (or associated with an inspection finding of Violation associated.with Significance Determination Process evaluates as having low to moderate, or greater, safety that the Significance safety significance significance (i.e.,

white, yellow, or red); or any order based upon upon aa violation.

violation.

66

Subject to the oversight and direction ofOE, Subject of OE, and with the approval of the Deputy Deputy Executive Director, where necessary, the regional offices normally issue Notices of Violation and proposed civil penalties. However, subject to the same oversight oversight as the regional offices, the Office of Nuclear Reactor Reactor Regulation (NRR) and the Office of Nuclear Material Safety Safety and Safeguards (NMSS) may also issue Notices of Violation and proposed civil penalties for certain Safeguards activities. Enforcement Enforcement orders are normally issued by the Deputy Executive Director or the Executive Director Director, OE. However, orders may also be issued by the EDO, especially Director,OE. especially those involving the more significant matters. The Directors ofNRR of NRR and NMSS have also been delegateddelegated authority to issue orders, but it is expected that normal use of this authority by NRR and NMSS will be confined to actions not associated associated with compliance issues. The Chief Financial Officer has been Financial Officer been delegated the authority authority to issue orders where where licensees violate violate Commission regulations by by nonpayment of license and inspection fees.

oflicense In recognition that the regulation regulation of ofnuc1ear activities in many cases does not lend itself to nuclear activities a mechanistic mechanistic treatment, judgment and discretion discretion must be exercised in determining determining the severity levels of the violations enforcement sanctions, including violations and the appropriate enforcement including the decision to issue a Notice of Violation, or to propose or impose a civil penalty and the amount of this penalty, after after considering the general principles considering principles of this statement statement of policy and the significance significance of the violations and the surrounding circumstances.

surrounding circumstances.

Unless Commission consultation or notification is required required by this policy, the NRC staff staff may depart, where warranted warranted in the public's interest, from this policy as provided in Section VII, "Exercise "Exercise of Discretion."

The Commission will be provided written notification for the following situations:

(1) All enforcement enforcement actions involving involving civil penalties penalties or orders; (2)

(2) The first time that discretion is exercised exercised for a plant that meets the criteria criteria of of Section Section VII.B.2; (3)

(3) (Where significance of the issue) when (Where appropriate, based on the uniqueness or signi-qcance discretion is exercised exercised for violations that meet the criteria of Section VII.B.6; VII.B.6; and (4) All Notices of Enforcement Enforcement Discretion (NOEDs) issued involving natural events, such as severe weather conditions.

conditions.

The Commission will be consulted consulted prior to taking action action in the following situations (unless the urgency of the situation situation dictates immediate action):

(1)

(1 ) An action action affecting a licensee's operation that requires requires balancing the public health and safety or common defense and security implications implications of not operating against the potential potential radiological radiological or other hazards associated associated with continued operation (cases(cases involving involving severe weather 7

other natural or other natural phenomena phenomena maymay be addressed addressed by by the NRC NRC staff staff without without prior prior Commission Commission consultation consultation in in accordance accordance with with Section Section VILC);

VII.C);

(2)

(2) Proposals to impose Proposals impose a civilcivil penalty penalty for a single violation or single violation or problem problem that is greater greater than 33 times Severity Level times the Severity Level I value shown in in Table Table 1A of licensee; 1A for that class oflicensee; (3)

(3) Any proposed enforcement action that involves proposed enforcement involves a Severity Severity Level Level I violation; (4) believes warrants Any action the EDO believes warrants Commission Commission involvement; involvement; (5) Any proposed (5) enforcement case proposed enforcement case involving Office of involving an Office Investigations (01) report ofInvestigations where where the NRC staff (other (other than the 01 staff) stafi) does not arrive at the same conclusions those in conclusions as those the 01 report concerning issues of intent report concerning intent if the Director Director of 01 concludes that Commission 01 concludes Commission consultation warranted; and consultation is warranted; and (6)

(6) Any proposed enforcement action on which the Commission proposed enforcement Commission asks to be consulted.

SIGNIFICANCE OF VIOLATIONS IV. SIGNIFICANCE VIOLATIONS requirements44 have varying degrees Regulatory requirements Regulatory degrees of safety, safeguards, safeguards, or environmental environmental significance. Therefore, significance. Therefore, the relative importance importance or significance significance of each violation violation is assessed as first step in the enforcement process.

the frrst A. Assessing Significance Significance In assessing the significance noncompliance, the NRC considers four specific significance of a noncompliance, specific issues:

(1) actual safety consequences; consequences; (2) safety consequences, (2) potential safety consequences, including the consideration of consideration of risk information; (3) impacting the NRC's ability to perform (3) potential for impacting regulatory function; perform its regulatory and (4) any willful aspects of the violation.

commercial nuclear power plants, the NRC relies on For certain types of violations at commercial on information information from the Reactor Oversight Process's Significance Significance Determination Process (SDP).

The SDP is used to evaluate significance of inspection findings to evaluate the actual and potential safety significance provide a risk-informed framework framework for discussing and communicating the significance significance of of inspection findings. Violations associated with fmdings inspection fmdings. findings evaluated evaluated through the SDP are addressed in Section IV.A.5. Violations at commercial nuclear power addressed power plants that are associated associated with inspection findings that cannotcannot be evaluated through the SDP (i.e., (i.e., violations that may impact the NRC's ability for oversight of licensed activities and violations that involve willfulness, including discrimination) are evaluated accordance with the guidance in evaluated in accordance IV.A. 1 through IV.A.4 and Section IV.B. Violations that are associated with inspection Sections IY.A.l inspection 4 The term "requirement" 4The term "requirement" as as used used in in this this policy policy means means aa legally legally binding binding requirement such as a statute, regulation, license condition, technical specification, specification, or order.

8

findings with actual consequences are evaluated actual consequences accordance with the guidance in evaluated in accordance Section IY.A.5.c.

IV.A.5.c.

1. Actual Safety Consequences.

Consequences. In evaluating actual safety consequences, consequences, the NRC considers issues such as actual onsite or offsite releases of radiation, onsite or offsite radiation radiation exposures, accidental criticalities, core damage, loss of significant safety barriers, barriers, loss of control control of radioactive radioactive material material or radiological radiological emergencies. (See Section Section IV.A.5.c IV.A.5.c for guidance guidance on on violations that are associated associated with SDP findings with actual consequences.)

consequences.)

2. Potential Potential Safety Consequences.

Consequences. In evaluating evaluating potential potential safety consequences, the NRC considers the realistic realistic likelihood of affecting safety, i.e., the existence existence of credible scenarios with potentially consequences. The NRC will use risk information potentially significant actual consequences. information wherever wherever possible possible in assessing significance and assigning assigning severity levels. A higher severity may be warranted for violations that have greater risk significance significance and a lower severity level hive 1may be appropriate for issues that have low risk significance. Duration is an appropriate appropriate appropriate consideration consideration in in assessing the significance significance of violations.

3. Impacting Regulatory Process.

Impacting the Regulatory Process. The NRC considers the safety safety implications of of noncompliances noncompliances that may impact impact the NRC's ability to carry out it statutory statutory mission.

Noncompliances Noncompliances may be significant because because they may challenge envelope upon challenge the regulatory envelope upon which certain certain activities were licensed. These These types of violations include include failures such as:

failures to provide provide complete and accurate information, failures to receive receive prior NRC approval for for changes in licensed activities, failures to notii)r notify NRC of changes changes in licensed licensed activities, activities, failure to perform 10 CFR 50.59 analyses, reporting failures, etc., etc., Even inadvertent reporting failures are important because many of the surveillance, surveillance, quality control, and auditing systems on which both the NRC and its licensees rely in order to monitor compliancecompliance with safety standards are based primarily on complete, accurate, recordkeeping and reporting. The existence of a complete, accurate, and timely recordkeeping regulatory automatically mean regulatory process violation does not automatically mean that the issue is safety significant.

significant. InIn determining the significance significance of a violation, the NRC will consider appropriateappropriate factors for the particular regulatory process violation. These factors may include: the significance significance of the underlying issue, whether whether the failure actually impeded or influenced influenced regulatory action, the level of individuals involved in the failure and the reasonableness reasonableness of the failure given their positionposition and training, and whether invalidates the licensing whether the failure invalidates consider for licensing basis. Factors to consider for failures to provide provide complete and accurate information are addressed addressed in Section IX of this policy.

Unless otherwise otherwise categorized categorized in the Supplements Supplements to this policy statement, statement, the severity severity level of a violation involving involving the failure to make a required report to the NRC will be based upon upon the significance significance of and the circumstances circumstances surrounding surrounding the matter matter that should have been reported.

However, the severity level of an untimely report, in contrast to no report, may be reduced reduced circumstances surrounding depending on the circumstances surrounding the matter. A licensee will not normally normally be cited for for a failure to report a condition or event unless the licensee was actually actually aware aware of the condition or or event that it failed to report. A licensee will, on the other hand, normally be cited for a failure to 9

report a condition or event if the licensee licensee knew of the information to be reported, but did not recognize recognize that it was required to make a report.

4.

4. Willfulness. Willful violations are by definition definition of particular particular concern to the Commission because its regulatory regulatory program is based on licensees licensees and their contractors, employees, and agents acting with integrity and communicating communicating with candor. Willful violations cannot be tolerated by either the Commission or a licensee. Therefore, a violation may be considered more significant than the underlying noncompliance noncompliance if it includes indications of of willfulness. The term "willfulness" as used in this policy embraces a spectrum of violations ranging from deliberate deliberate intent to violate or falsify to and including including careless disregard for for requirements. Willfulness does not include acts which do not rise to the level of careless disregard, e.g., negligence or inadvertent inadvertent clerical clerical errors in a document submitted to the NRC. In In determining the significance significance of a violation involving willfulness, consideration consideration will be given to such factors as the position and responsibilities responsibilities of the person involvedinvolved in the violation (e.g., (e.g.,

official5 or non-supervisory licensee officials non-supervisory employee), the significance significance of any underlying underlying violation, the intent of the violator (i.e.,

(i.e., careless careless disregard disregard or deliberateness),

deliberateness), and the economic economic or other other advantage, if any, gained as a result of the violation. The relative weight given to each of these factors in arriving at the significance assessmentassessment will be dependent dependent on the circumstances circumstances of the violation. However, if a licenseelicensee refuses to correct a minor violation within a reasonable reasonable time such that it willfully continues, continues, the violation should be considered at least more than minor.

Licensees Licensees are expected to take significant significant remedial action in responding responding to willful violations commensurate with the circumstances circumstances such that it demonstrates demonstrates the seriousness seriousness of the violation violation creating a deterrent thereby creating deterrent effect within the licensee's licensee's organization.

5. Significance Significance Determination Determination Process.Process. The Reactor Reactor Oversight Process uses a Determination Process Significance Determination Process (SDP) significance of most (SDP) to determine the safety significance inspection inspection findings identified identified at commercial nuclear power power plants. Depending Depending on their significance, inspection significance, inspection findings are assigned colors of green, white, yellow, or red. The Reactor Reactor Oversight Process uses an Agency Agency Action Matrix to determine the appropriate agency agency response.

If violations that are more than minor are associated with these inspection Ifviolations findings, they will be inspection [mdings, documented and mayor may or may not be cited depending on the safety significance. significance. These violations are not normally assigned assigned severity levels, nor are they normally normally subject to civil penalties.

NOTE: Violations associated associated with inspection findings that are not evaluated evaluated through through the SDP SDP will be assigned severity levels in accordance accordance with Section IV.B and will be subject subject to civil penalties penalties in accordance with Section VI.C.

Section Vl.C.

5 The term 5The term "licensee "licensee official" official" as used in this policy statement statement means a first-line supervisor supervisor or above, above, a licensed licensed individual, a radiation safety officer, safety officer, or an authorized user of licensed licensed material whether or not listed on a license.

Notwithstanding Notwithstanding an individual's job title, severity severity level categorization categorization for willful acts involving individuals who can can be considered considered licensee officials officials will consider several factors, including including the position position of the individual relative to the licensee's organizational organizational structure and the individual's responsibilities responsibilities relative to the oversight oversight of licensed activities oflicensed and to the use of licensed material.

material.

10

a. Violations Associated Associated with Findings of Very Low Safety Significance Safety Significance Violations associated associated with findings that the SDP evaluates as having very low safety safety significance (i.e.,

significance (i.e., green) normally be described in inspection reports as Non-Cited will nonnally Violations (NCVs). The finding will be categorized categorized by the assessment process within the licensee response response band. However, a Notice of Violation (NOV) will be issued if the issue meets one of the three applicable exceptions in Section VI.A.l.

applicable exceptions VI.A. 1. The Commission recognizes recognizes that violations exist below this category category that are of minimal safety or environmental environmental significance.

While licensees must correctcorrect these minor violations, they don't normally documentation nonnally warrant documentation in inspection reports and do not warrantwarrant enforcement enforcement action. To the extent such violations are described, they will be noted as violations of minor significance significance that are not subject to enforcement action.

enforcement

b. Violations Associated Associated with Findings of Low to Moderate, Greater Safety Moderate, or Greater Safety Significance Significance Violations associated associated with findings that the SDP evaluates evaluates as having having low to moderate moderate safety significance significance (i.e.,

(i.e., white), substantial safety significance (yellow),

safety significance (yellow), or high safety safety significance (red) will be cited in an NOV requiring a written response significance response unless sufficient sufficient information infonnation is already on the docket. The finding will be assigned a color related to its significance significance for use by the assessment assessment process. The Commission reserves reserves the use of discretion discretion for particularly particularly significant significant violations (e.g. an accidental accidental criticality) to assess civil penalties inin accordance accordance with Section 234 of the Atomic Energy Act of 1954, as amended.

c. Violations Associated with Actual Consequences Consequences Violations Violations that involve actual consequences consequences such as an overexposure overexposure to the public or or plant personnel above regulatory regulatory limits, failure to make the required notifications that impact the ability of Federal, State and local agencies agencies to respond to an actual emergency preparedness (site emergency preparedness (site area or general emergency), transportation event, or a substantial emergency), transportation substantial release of radioactive radioactive material, will be assigned severity levels and will be subject to civil penalties.

B. Assigning Severity Level Level For purposes of determining the appropriate appropriate enforcement enforcement action, violations (except (except the majority of those associated associated with findings evaluated evaluated though the SDP) are normally nonnally categorized in categorized terms tenns of four levels of severity to show their relative importance importance or significance significance within each of of the following eight activity areas:

I.

1. Reactor Operations; I].

II. Facility Facility Construction; 111.

Ill. Safeguards; IV. Health Physics; 11 11

V. Transportation; Transportation; VI. Fuel Cycle and Materials Fuel Materials Operations; VII.

VII. Miscellaneous Matters; Miscellaneous Matters; and VIII. Emergency Preparedness.

Emergency Preparedness.

Licensed Licensed activities activities will will be placed placed inin the activity area area most most suitable suitable in in light of of the the particular violation particular violation involved, involved, including including activities not directly directly covered covered by one of the listed areas, e.g.,

e.g., export license activities. Within each each activity activity area, Severity Level I has been assigned assigned to to violations that are are the most significant significant and Severity Severity Level IV violations violations are the least significant.

Severity Severity Level Level I and II violations violations are ofof very very significant regulatory concern.66 In general, significant regulatory general, violations that are are included included in these these severity severity categories categories involve involve actual actual or high potential potential consequences consequences on on public public health health and safety. Severity Severity Level Level III violations violations are are cause cause for significant significant regulatory concern. Severity Severity Level IV violations are less seriousserious but but are of more than minor minor concern. Violations at Severity Severity Level Level IV involve noncompliance noncompliance with NRC requirements requirements that are not considered significant based significant based on risk. This should not be misunderstood to imply imply that that Severity Severity Level IV issues issues have have no risk significance.

The Commission Commission recognizes recognizes that there are other violations violations of minor safety or or environmental concern environmental concern that are below the level of significance of Severity significance Severity Level IV IV violations.

violations.

licensees must While licensees must correct correct these minor minor violations, violations, they don't normally warrant warrant documentation documentation in inspection reports or inspection inspection records and do not warrant enforcement enforcement action. To the extent such violations are described, they will be noted as violations of minor significance significance that are not not subject to enforcement enforcement action.

Comparisons of significance significance between activity areas are inappropriate.

inappropriate. For example, example, the immediacy immediacy of any hazard hazard to the public associated with Severity Severity Level I violations in Reactor Reactor Operations Operations is not directly directly comparable comparable to that associated with Severity Level I violations in Facility Construction.

Supplements I through VIII Supplements VIn provide provide examples and serve as guidance guidance in determining the appropriate severity level for violations in each of the eight activity areas. However, the appropriate examples are neither exhaustive exhaustive nor controlling. In addition, these examples examples do not create new new requirements. Each is designed designed to illustrate illustrate the significance that the NRC placesplaces on a particular particular type of violation of NRC requirements. Each of the examples in the supplements is predicated predicated on a violation of a regulatory requirement.

requirement.

The NRC reviews each case being considered considered for enforcement action on its own merits to ensure that the severity of a violation is characterized characterized at the level best suited to the significance significance of the particular violation.

6 Regulatory concern pertains to primary NRC regulatory responsibilities, i.e., safety, safeguards, and the 6 Regulatory concern pertains to primary NRC regulatory responsibilities, i. e., safety, safeguards, and the environment.

12 12

v.

V. PREDECISIONAL PREDECISIONAL ENFORCEMENT ENFORCEMENT CONFERENCES CONFERENCES When the NRC leamslearns of a potential violation for which escalated enforcement which escalated enforcement action action appears to be warranted, or recurring nonconformance nonconformance on the part of a contractor, the NRC may provide an opportunity for a predecisional predecisional enforcement enforcement conference conference with the licensee, contractor, or other person person before taking enforcement enforcement action. The purpose of the predecisional predecisional enforcement enforcement conference conference is to obtain information that will assist the NRC in determining the appropriate appropriate enforcement enforcement action, such as: (1) a common understanding understanding of facts, root causes, and missedmissed opportunities opportunities associated associated with the apparent apparent violations; (2) (2) a common understanding understanding of corrective actions taken or planned; planned; and (3)

(3) a common understanding significance of issues and the understanding of the significance need for lasting comprehensive comprehensive corrective corrective action.

The NRC may conduct conduct Regulatory Conferences (in lieu of predecisional Regulatory Conferences predecisional enforcement enforcement conferences) to discuss the significance of conferences) findings evaluated offindings evaluated by the Reactor Oversight Process's Process's SDP when apparent violations are associated with potentially significant significant findings. The purpose of Regulatory Conferences Conferences is to get information information from licensees licensees on the significance of findings significance offindings evaluated evaluated through the SDP whether or not violations are involved. Because the significance significance assessment assessment from the SDP determines whether whether or not escalated enforcement action escalated enforcement action will will be issued issued (i.e., a Notice of Violation associated (i.e., Notice ofYiolation associated with a white, yellow, or red SDP finding), a subsequent subsequent predecisional enforcement predecisional enforcement conference conference is not normally necessary.

If the NRC concludes that it has sufficient information information to make an informed enforcement decision involving involving a licensee, contractor, or vendor, a predecisional enforcement conference pre decisional enforcement conference will not normally be held. If a predecisional enforcement conference predecisional enforcement conference is not held, the licensee may be given an opportunity opportunity to respond to a documented apparent apparent violation (including its root causes and a description of planned planned or implemented corrective actions) before the NRC takes implemented corrective enforcement action. However, if enforcement the NRC has sufficient information to conclude ifthe conclude that a civil penalty penalty is not warranted, it may proceed to issue an enforcement enforcement action without first obtaining the licensee's licensee's response to the documented apparent apparent violation.

The NRC will normally provide an opportunity opportunity for an individual to address address apparent apparent violations before the NRC takes escalated enforcement action. Whether an individual will be escalated enforcement provided opportunity for a predecisional provided an opportunity predecisional enforcement conference or an opportunity enforcement conference opportunity to address address an apparent violation in writing writing will depend on the circumstances circumstances of the case, including the severity of the issue, the significance of the action the NRC is contemplating, and whether whether the individual has already had an opportunity to address the issue (e.g., (e.g., an Office of Investigation or Department of Labor hearing).

a Department During the predecisional predecisional enforcement conference, the licensee, contractor, or other enforcement conference, persons will be given an opportunity to provide information persons information consistent with the purpose of the conference, including an explanation conference, corrective actions (if any) that explanation to the NRC of the immediate corrective were taken following identification of the potential potential violation violation or nonconformance nonconformance and the comprehensive actions that were taken or will be taken to prevent recurrence.

long-term comprehensive 13

Licensees, persons will be told when a meeting is a predecisional Licensees, contractors, or other persons predecisional enforcement conference.

enforcement A predecisional predecisional enforcement conference is a meeting between enforcement conference between the NRC and the licensee.

Conferences are normally held in the regional Conferences regional offices offices and are normally normally open to public public observation. Predecisional Predecisional enforcement conferences will not normally be open to the public if enforcement conferences if the enforcement enforcement action being contemplated:

though not taken against an (1) Would be taken against an individual, or if the action, though an individual, turns on whether an individual has committed committed wrongdoing; (2)

(2) Involves personnel failures where the NRC has requested significant personnel Involves significant requested that the individual(s) individual(s) involved conference; involved be present at the conference; (3) Is based on the findings of an NRC (3) NRC Office of Investigations report that has not been oflnvestigations disclosed; or publicly disclosed; or (4) Involves safeguards information, Privacy Act information, or information Involves safeguards information which could be considered considered proprietary; conferences will not normally be open to the public if:

In addition, conferences if:

(5)

(5) The conference conference involves involves medical misadministrations overexposures and the misadministrations or overexposures conference cannot be conducted without disclosing the exposed conference cannot exposed individual's name; or or (6) The conference (6) conducted by telephone or the conference conference will be conducted conference will be conducted at a relatively small licensee's facility.

Notwithstanding meeting any of these criteria, a predecisional Notwithstanding predecisional enforcement enforcement conference conference may still be open ifthe conference involves issues related to an ongoing adjudicatory if the conference adjudicatory proceeding with one or more interveners or where evidentiary basis for the conference where the evidentiary conference is a matter of of public record, such as an adjudicatory adjudicatory decision Department of Labor. In addition, decision by the Department notwithstanding the normal criteria for opening or closing predecisional enforcement notwithstanding enforcement conferences, conferences may either conferences, conferences either be open or closed to the public, with the approval approval of the Executive Executive Director for Operations, after balancing the benefit benefit of the public's observation observation against against the potential impact on the agency's decision-making decision-making process process in a particular case.

The NRC will notifY notify the licensee that the predecisional predecisional enforcement conference enforcement conference .

will .be open to public observation. Consistent Consistent with the agency's policy on open meetings (included (included on the NRC's Public Meeting Web site), the NRC intends to announce open conferences conferences normally at calendar days in advance least 10 calendar advance of conferences.

conferences. Conferences Conferences will be announced announced on the Internet Internet at the NRC Office of Enforcement's Enforcement's homepage (www.nrc.gov/OE)

(www.nrc.gov/OE) and on the Public Meeting Meeting (www.nrc.gov/NRC/PUBLIC/meet.html). Individuals who do not have Internet access Web site (www.nrc.gov/NRC/PUBLIC/meet.html). access 14

may get assistance on scheduled conferences conferences by contacting the NRC staff at the Public Document Document Room, by calling toll-free 1-800-397-4209. In addition, the NRC will normally issue a press toll-free 1-800-397-4209.

release and notify appropriate appropriate State liaison officers predecisional enforcement officers that a predecisional enforcement conference conference has been scheduled scheduled and that it is open to public observation.

The public attending open predecisional predecisional enforcement enforcement conferences may observe observe but may not participate in the conference. The purpose of conducting conducting open conferences conferences is not to maximize public attendance, but rather maximize rather to provide the public with opportunities opportunities to be informed of of NRC activities consistent with the NRC's ability to exercise exercise its regulatory regulatory and safety responsibilities. Therefore, members of the public will be allowed access to the NRC regional offices to attend open enforcement enforcement conferences conferences in accordance accordance with the "Standard OperatingOperating Procedures Procedures For Providing Security Support For NRC Hearings and Meetings," Meetings," published published November 1, 1, 1991 (56 FR 56251). These procedures procedures provide that visitors may be subject subject to personnel personnel screening, screening, that signs, banners, posters, etc., etc., not larger than 18" 18" be permitted, and that disruptive persons may be removed. The open conference conference will be terminated if disruption disruption interferes interferes with a successful successful conference.

conference. NRC's Predecisional Predecisional Enforcement Enforcement Conferences Conferences (whether (whether open or closed) normally will be held at the NRC's regional offices or in NRC Headquarters Headquarters Offices Offices and not in the vicinity of the licensee's facility.

For a case in which an NRC Office ofInvestigations of Investigations (01) report finds that discrimination discrimination as defined defined under 10 CFR 50.7 (or similar provisions in Parts 30, 40, 60, 70, or 72) has occurred, the 01 report may be made public, subject to withholding withholding certain (i.e., after certain information (i.e., after appropriate appropriate redaction),

redaction), in which case the associated predecisional enforcement associated predecisional conference will enforcement conference normally be open to public observation. In a predecisional predecisional enforcement conference where a enforcement conference particular particular individual is being considered considered potentially potentially responsible responsible for the discrimination, the conference conference will remain closed. In either case (i.e., conference is open or closed), the (i.e., whether the conference employee or former employee who was the subject employee subject of the alleged discrimination (hereafter discrimination (hereafter referred referred to as "complainant")

"complainant") will normally be provided an opportunity to participate in the predecisional enforcement conference predecisional enforcement conference with the licensee/employer. This participationparticipation will normally normally be in the form of a complainant complainant statement statement and comment comment on the licensee's presentation, presentation, followed in tum turn by an opportunity opportunity for the licensee licensee to respond to the complainant's presentation.

In cases where where the complainant complainant is unable to attend in person, arrangements arrangements will be made for the complainant's complainant's participation by telephone telephone or an opportunity opportunity given for the complainant complainant to submit a written response response to the licensee's licensee's presentation. If the licensee chooses to forego an enforcement enforcement conference conference and, instead, responds to the NRC's findings in writing, the complainant complainant will be provided provided the opportunity to submit written comments on the licensee's response. For cases cases involving potential potential discrimination discrimination by a contractor, any associated predecisional enforcement predecisional enforcement conference with the contractor conference contractor would be handled similarly. These arrangementsarrangements for complainant complainant participation participation in the predecisional predecisional enforcement enforcement conference conference are not to be conducted conducted or viewed in any respect as an adjudicatory adjudicatory hearing. The purpose of the complainant's ofthe complainant's participation is to to provide provide information to the NRC to assist it in its enforcement enforcement deliberations.

15 15

A predecisional predecisional enforcement conference conference may not need to be held in cases where there is a full adjudicatory adjudicatory record before the Department Department of Labor. If a conference conference is held in such cases, cases, generally generally the conference conference will focus on the licensee's corrective corrective action. As with discrimination discrimination cases based on 01 investigations, investigations, the complainant complainant may be allowed to participate.

Members of the public attending Members attending open predecisional predecisional enforcement enforcement conferences conferences will be reminded that (1) the apparent violations discussed at predecisional predecisional enforcement conferences are enforcement conferences subject to further review and may be subject to change change prior to any resulting enforcement enforcement action and (2)

(2) the statements statements of views or expressions of opinion made by NRC employees employees at predecisional enforcement conferences, predecisional enforcement conferences, or the lack thereof, thereof, are not intended intended to represent represent final determinations or beliefs.

determinations When needed to protect protect the public health and safety or common common defense defense and security, escalated enforcement action, such as the issuance of an immediately escalated enforcement immediately effective effective order, will be taken before the conference.

conference. In these cases, a conference conference may be held after the escalated escalated enforcement action is taken.

enforcement VI. DISPOSITION OF VIOLATIONS VIOLATIONS This section describes describes the various ways the NRC can disposition violations. The manner manner in which a violation violation is dispositioned is intended intended to reflect the seriousness seriousness of the violation and the circumstances involved. As previously circumstances previously stated, minor violations are not the subject of subject of enforcement enforcement action. While While licensees must correct these violations, they don't normally warrant documentation in inspection reports or inspection records. Other violations are documented documentation documented and may be dispositioned as Non-Cited Non-Cited Violations, Violations, cited in Notices of Violation, or issued in conjunction with civil penalties or various various types of orders. The NRC may also choose to exercise discretion and refrain refrain from issuing enforcement enforcement action. (See (See Section VII.B, "Mitigation "Mitigation of of Enforcement Sanctions.")

Enforcement Sanctions.") As discussed further in Section VI.E, related administrative administrative actions such as Notices of Nonconformance, Nonconformance, Notices of Deviation, Confirmatory Confirmatory Action Letters, Letters of Reprimand, and Demands for Information are used to supplement the enforcement enforcement program.

In determining the appropriate appropriate regulatory response, the NRC will consider enforcementenforcement actions taken by other Federal Federal or State State regulatory bodies having concurrent jurisdiction, such as in in transportation transportation matters.

A.

A. Non-Cited Non-Cited Violation Violation (NCV)

A Non-Cited A Non-Cited Violation Violation (NCV)

(NCV) is the term is the term used to describe a method for dispositioning a Severity Severity Level IV violation violation or a violation associated with a finding that the Reactor Reactor Oversight Oversight Process's SDP evaluates as having very low safety significance significance (i.e.,

(i.e., green). These issues are documented as violations in inspection reports (or inspection records for some materials materials licensees) licensees) to establish public public records records of the violations, but are not cited in Notices of Violation Violation which normally require written responses from licensees (see Section VI.B below).

Dispositioning Dispositioning violations in this manner does not eliminate the NRC's emphasis emphasis on compliance compliance 16 16

importance of maintaining with requirements nor the importance maintaining safety. Licensees are still responsible responsible for maintaining safety and compliance compliance and must take steps to address corrective corrective actions for these violations. While licensees are not required to provide written responses to NCVs, this approach allows licensees to dispute violations describeddescribed as NCVs. The following sections describe the circumstances under which a violation may circumstances mayor or may not be dispositioned dispositioned as an NCV.

1. Power Reactor

1. Power Reactor Licensees Licensees Severity Level IV violations and violations associated with green SDP findings are normally dispositioned dispositioned as NCVs. Violations dispositioned as NCVs will be described in inspection reports, although the NRC will close these violations based on their being entered inspection entered into the licensee's corrective corrective action program. At the time a violation is closed in an inspection report, the licensee may not have completed its corrective corrective actions or begun begun the process identify the process to identify root cause and develop action to prevent recurrence. recurrence. Licensee Licensee actions will be taken commensurate with the established priorities and processes of the licensee's corrective action commensurate program. The NRC inspection inspection program program will provide provide an assessment assessment of the effectiveness of the corrective action program. In addition to documentation in inspection reports, violations will be corrective entered into the Plant Issues MatrixMatrix (PIM).

(PIM). Because Because the NRC will not normally obtain a written licensees describing actions taken to restore compliance response from licensees compliance and prevent recurrence of prevent recurrence of these violations, this enforcement enforcement approach places greater greater NRC reliance on licensee licensee corrective corrective action programs. AnyoneAny one of the following circumstances circumstances will result in consideration of an NOV requiring a formal written response from a licensee.

a. The licensee failed to restore compliance compliance within a reasonable time after a violation was identified.

b. The licensee did not place the violation into a corrective corrective action program to address recurrence.

c. The violation is repetitive 7 as a result of inadequate corrective

c. corrective action, and was identified by the NRC. NOTE: This exception does not apply to violations associated associated with green SDP findings.

d. The violation was willful. Notwithstanding willfulness, willfulness, an NCV may still be appropriate if:

appropriate if:

7 7A violation is considered "repetitive" "repetitive" if it could reasonably expected to have been prevented reasonably be expected licensee's prevented by the licensee's corrective action for a previous corrective previous violation or a previous licensee licensee finding that occurred within the past 22 years of the inspection inspection at issue, or the period within the last two inspections, whichever is longer.

17 17

(1) The The licensee licensee identified identified the violation violation and and the information information concerning concerning the violation, if not required to be reported, not required reported, was promptly promptly provided provided to appropriate appropriate NRC NRC personnel, personnel, such as as a resident resident inspector inspector or regional regional branch branch chief; (2)

(2) The violation violation involved involved the acts of a low-level low-level individual (and(and not a licensee licensee official official as defined in Section Section IV.A);

(3) The violation appears to be the isolated isolated action action of the employee employee without without management management involvement involvement and the violation violation was not caused caused by lack of management management oversight oversight as evidenced evidenced by eithereither a history of isolated willful willful violations violations or a lack of adequate adequate audits or or supervision supervision of employees; employees; and (4) Significant Significant remedial remedial action commensurate commensurate with the circumstances circumstances was takentaken by by the licensee licensee such such that it demonstrated seriousness of the violation to other employees demonstrated the seriousness employees and contractors, contractors, thereby thereby creating creating a deterrent effect within the licensee's licensee's organization.

The approval approval of the Director, Office of Enforcement, Enforcement, with consultation consultation with the Deputy Deputy Executive Executive Director as warranted, is required required for dispositioning willful violations as NCVs.

2.

2. - 7. [Reserved]

8. All Other Licensees Licensees Severity Level IV IV violations that are dispositioned as NCVs will be described in inspection reports (or inspection inspection records for some materials materials licensees) licensees) and will include a brief brief description of the corrective description ofthe corrective action the licensee has either taken or planned to take. Any one of Anyone of the following circumstances circumstances will result in consideration consideration of an NOV requiring a formal written written response from a licensee.

a. The licensee failed to identify the violation;'

violation;8

b. The licensee did not correct or commit to correct the violation within a reasonable time by specific corrective action committed to by the end of the inspection, including immediate corrective action and comprehensive comprehensive corrective corrective action to prevent recurrence; and

c. The violation is repetitive as a result of inadequate corrective corrective action; 8'An An NOV is warranted when a licensee identifies a violation as aa result of an event where the root cause of the event is obvious or the licensee had prior opportunity event opportunity to identify the problem but failed to take action that would event. Disposition as an NCV may be warranted if the licensee demonstrated initiative in have prevented the event.

identifying the violation's root cause.

18 18

d. The violation was willful. Notwithstanding Notwithstanding willfulness, willfulness, an NCV may still be appropriate appropriate if it meets the criteria in Section VI.A. 1.d.

Section VI.A.l.d.

The approval of the Director, Office of Enforcement, Enforcement, with consultation with the Deputy Executive Director as warranted, is required for dispositioning willful violations as NCVs.

Executive B. Notice Notice of Violation A Notice of Violation is a written notice setting forth one or more violations of a legally legally binding requirement. The Notice of ViolationViolation normally requires the recipient to provide provide a statement describing (1) the reasons for the violation or, if contested, the basis for written statement for disputing the violation; violation; (2)(2) corrective corrective steps that have been taken and the results achieved; (3) corrective corrective steps that will be taken to prevent prevent recurrence; recurrence; and (4) the date when full compliance compliance will be achieved. The NRC may waive all or portions of a written response response to the extent that relevant information information has already been provided to the NRC in writing or documented documented in an NRC inspection inspection report or inspection inspection record. The NRC may require responses to Notices of of Violation Violation to be under oath. Normally, responses under oath will be required required only in connection connection with Severity Level Levell, I, II, II, or III violations; violations associated associated with findings that the SDP evaluates evaluates as having low to moderate, or greater safety significance (i.e., white, yellow, or red); or significance (i.e., or orders.

Issuance Issuance of ofaa Notice of Violation is normally the only enforcement enforcement action taken for Severity Level Levell,I, II, and III violations, except in cases where the criteria criteria for issuance of civil penalties penalties and orders, as set forth in Sections VI.C and VI.D, respectively, are met.

C. Civil Penalty Penalty A civil penalty penalty is a monetary penalty that may be imposed for violation of of(1)

(1) certain specified licensing provisions of the Atomic Energy Act or supplementary supplementary NRC rules or orders; (2)

(2) any requirement requirement for which a license may be revoked; revoked; or (3) reporting reporting requirements under under section 206 of the Energy Reorganization Reorganization Act. Civil penalties are designed to deter future violations both by the involved involved licensee licensee and other licensees conducting conducting similar activities. Civil penalties penalties also emphasize emphasize the need for licensees to identify violations and take prompt prompt comprehensive comprehensive corrective action.

Civil penalties are normally normally assessed for Severity Levell Level I and II violations and knowing and conscious violations of the reporting requirements of section 206 of the Energy Reorganization Act. Civil penalties are considered Reorganization considered for Severity Level III violations.

Civil penalties are also considered for violations associated with inspection inspection findings evaluated through the Reactor Reactor Oversight Oversight Process's SDP that involved involved actual consequences, such such overexposure to the public or plant personnel as an overexposure personnel above regulatory limits, failure to make the 19 19

required required notifications notifications that that impact impact the ability ability of Federal, Federal, State and local local agencies agencies to .respond respond to anan actual emergency emergency preparedness preparedness event event (site general emergency),

(site area or general transportation event, or a emergency), transportation substantial release substantial (Civil penalties radioactive material. (Civil release of radioactive penalties are are not proposed proposed for violations violati9ns associated associated with low to moderate, or greater safety significant greater safety significant findings absent absent actual consequences.

consequences.))

Civil penalties penalties are used to encourage prompt and identification and prompt encourage prompt identification correction of violations, to emphasize comprehensive correction comprehensive compliance in aa manner emphasize compliance manner that deters deters future violations, violations, and to serve licensees' attention serve to focus licensees' attention on on significant significant violations.

Although management direct or involvement, direct management involvement, or indirect, indirect, in a violation may lead to an an lack of management increase in the civil penalty, the lack increase management involvement involvement may not not be used to mitigate a to mitigate mitigation in the latter case could civil penalty. Allowing mitigation encourage the lack of management could encourage management involvement in licensed activities and a decrease in protection involvement protection of the public public health and safety.

1. Base Civil Penalty Penalty The NRC imposes different penalties for different different levels of penalties different severity severity level level violations and violations and different classes of contractors, and other persons. Violations that involve licensees, contractors, oflicensees, involve loss, abandonment, or improper transfer abandonment, transfer or disposal source or device are treated disposal of a sealed source treated separately, separately, regardless regardless of the use or the type of licensee.

licensee. Tables I IAA and l IBB show the base civil penalties for for various reactor, fuel cycle, and materials programs, and for the loss, abandonment abandonment or improper improper disposal of a sealed source or device. (Civil penalties transfer or disposal penalties issued to individuals individuals are case-by-case basis.)

determined on a case-by-case basis.) The structure of these tables generally takes into account structure ofthese account of the violation the gravity ofthe consideration and the ability to pay as a secondary violation as a primary consideration consideration. Generally, operations involving greater nuclear material inventories greater inventories and greater potential consequences consequences to the public and licensee employees employees receive higher civil penalties.

secondary factor of ability of various classes of licensees to pay the civil penalties, Regarding the secondary it is not the NRC's intention that the economic impact of a civil penalty be so severe severe that it puts a licensee out of business (orders, rather than civil penalties, penalties, are used when the intent is to suspendsuspend or terminate licensed activities) or adversely affects a licensee's ability to safely conduct licensed licensed activities. The deterrent effect of civil penalties is best served when the amounts ofthe of the penalties take into account a licensee's ability to pay. In determining the amount of civil penalties for for licensees for whom the tables do not reflect the ability to pay payor or the gravity of the violation, the necessary increases or decreases on a case-by-case basis. Normally, if a NRC will consider necessary licensee can demonstrate financial hardship, the NRC will consider payments over time; including interest, rather than reducing the amount of the civil penalty. However, where a licensee claims financial hardship, the licensee will normally be required to address why it has sufficient resources to safely conduct licensed activities and pay license and inspection fees.

20 20

TABLE 1A--BASE lA--BASE CIVIL PENALTIES

a. Power reactors and gaseous diffusion plants ...................... $130,000

. . . . . . . . . . . . . . . . . . .. $130,000

b. Fuel fabricators authorized authorized to possess Category Category I or II quantities quantities ofSNM of SNM .....................................

..................................... $65,000

c. Fuel fabricators, industrial processors, processors, I independent spent fuel and monitored and independent monitored retrievable retrievable storagestorage installations installations ................................

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. $32,500

d. Test reactors, mills and uranium conversion conversion facilities, contractors, waste disposal licensees, industrial radiographers, radiographers, and other large material .............................................

material users ............................................. $13,000

$13,000

e. Research reactors, academic, medical, or other small material users users22 ..................................

$6,500 ff. Loss, abandonment, abandonment, or improper improper transfer or disposal of a sealed sealed 3

source or device, regardless of the use or type regardless of the use or type of licensee: 3 of licensee:

1. Sources

1. Sources or devices with a total activity activity greater greater than 3.7 x 10' 4 MBq (1 Curie), excluding hydrogen-3 (tritium) ..........

3.7 x 10 MBq (1 Curie), excluding hydrogen-3 (tritium) .......... $50,000 $50,000

2. Other sources or devices containing containing the materials and quantities listed in 10 CFR 31.5(c)(13)(i) ..............................

31.5(c)(13)(i) .............................. $16,500

$16,500

3. Sources

3. Sources and devices not otherwise described ...............

described above ............... $6,500

'Large ILarge firms engaged in manufacturing 2

manufacturing or distribution of byproduct, source, or special nuclear material. material.

This applies 2This applies toto nonprofit nonprofit institutions institutions not not otherwise categorized in this table, mobile nuclear otherwise categorized nuclear services, services, nuclear pharmacies, pharmacies, and physician physician offices. .

3 These 3These base civil penalty amounts have been determined determined to be approximately three times the average average cost cost of disposal.

disposal. For specific cases, NRC may adjust these amounts to correspond to three times the actual expected expected cost of authorized authorized disposal.

TABLE TABLE IB--BASE lB--BASE CIVIL PENALTIES Severity Severity Level Level Base Civil Penalty Amount (Percent (Percent of amount listed in Table I1A)

I ..............................  ; ... 100%100%

II ................................... ......... 80%

80%

III .................................. 50%

500/0 21

2. Civil Penalty Assessment Civil Penalty Assessment In an effort effort to (1) emphasize emphasize the importance of adherence adherence to requirements requirements and (2) reinforce (2) reinforce prompt self-identification self-identification of problems and root causes causes and prompt and comprehensive comprehensive correction of violations, the NRC reviews each proposed civil penalty on its own merits and, after considering considering all relevant circumstances, circumstances, may adjust the base civil penalties penalties shown in Table lA shown 1A and IB lB for Severity Severity Level I, II, and III violations as described described below.

The civil penalty assessment assessment process considers considers four decisional points: (a) whether the licensee licensee has had any previous escalated enforcement action escalated enforcement action (regardless of the activity activity area) during the past 2 years or past 2 inspections, whichever is longer; (b) whether the licensee licensee should be given credit for actions related to identification; identification; (c)(c) whether the licensee's corrective actions are prompt and comprehensive; comprehensive; and (d) whether, in view of all the circumstances, circumstances, the matter in question requires the exercise of discretion. Although each of these decisional points may have several considerations for any given case, the outcome several associated considerations outcome of the assessment process for for each violation or problem, absent the exercise of discretion, is limited to one of the following following three results: no civil penalty, a base civil penalty, or a base civil penalty escalated by 100 100 percent. The flow chart presented presented below is a graphic representation representation of the civil penalty penalty assessment assessment process.

D Notice of YES YESViolation Violation "Civilof Notice Penalty D

ESCALATED PROCESS Nfor Nv 0

.NO Noticeo!

Violation ESCALATED PROCESS Severity Level 1,, denitcaio &

Severity and Levell, II IIIViolations SLIIi Base and III Violations D Civil NCorrecePenalty YES Penalty SDiscretioi D Notice of Violation NO 2 X Basa Discretion Civil Penalty

a. Initial Escalated Action

a. Initial Escalated Action When the NRC determines that a non-willful Severity Level II violation or problem has occurred, When and the the NRC licenseedetermines thatany has not had a non-willful Severity Level previous escalated actionsIII(regardless violation orofproblem the activity has occurred, and the licensee has not had any previous escalated actions (regardless of the activity area) during the past 2 years or 2 inspections, whichever is longer, the NRC will consider area) during the past 2 years or 2 inspections, whichever is longer, the NRC will consider whether the licensee's corrective action for the present violation or problem is reasonably prompt whether the licensee's corrective action for the present violation or problem is reasonably prompt and comprehensive (see the discussion under Section VI.C.2.c, below). Using 2 years as the and comprehensive (see the discussion under Section VI.C.2.c, below). Using 2 years as the basis for assessment is expected to cover most situations, but considering a slightly longer or basis shorter forperiod assessmentmight isbeexpected warranted to based cover onmost thesituations, circumstancesbut considering of a particulara slightly case. longer The starting or shorter point ofperiod mightshould this period be warranted based on be considered the the datecircumstances when the licensee of a particular was put on case. Theofstarting notice the need point to takeofcorrective this periodaction.

shouldFor be considered the date when a licensee-identified the licensee violation was put or an event, this on notice would be of when the theneed to take corrective action. For a licensee-identified violation or an event, this would be when the 22 22

licensee is aware that a problem or violation exists requiring licensee requiring corrective action. For an NRC-identified identified violation, the starting point would be when the NRC puts the licensee on notice, which which could be during the inspection, at the inspection could inspection exit meeting, meeting, or as part of post-inspection post-inspection communication.

communication.

If the corrective corrective action is judged to be prompt and comprehensive, a Notice of Violation Violation normally normally should be issued with no associated civil penalty. If the corrective corrective action is judged to to be less than prompt and comprehensive, comprehensive, the Notice of Violation normally should be issued issued with a base civil penalty.

b. Credit Creditfor for Actions Related to Identification Identification (1) If a Severity Severity Level I or II 1I violation or a willful Severity Level III violation has occurred--or occurred--or if, if, during the past 22 years or 2 inspections, whichever whichever is longer, the licensee licensee has been been issued at least one other escalated action--the escalated action--the civil penalty assessment should should normally normally consider identification in addition to corrective action (see the discussion under consider the factor of identification Section VI.C.2.c, VI.C.2.c, below). In these circumstances, circumstances, the NRC should consider consider whether whether the licensee licensee should be given credit for actions related to identification.

In each each case, the decision should be focused on identification identification of the problem requiring requiring corrective corrective action. In other words, although giving credit for Identification Identificationand Corrective CorrectiveAction should be separate separate decisions, the concept concept of Identification Identification presumes that the identifier recognizes identifier recognizes the existence existence of a problem, and understands that corrective action is needed. The decision on on Identification requires Identification requires considering considering all the circumstances circumstances of identification identification including:

(i)

(i) Whether the problem problem requiring corrective action was NRC-identified, requiring corrective NRC-identified, licensee-9 identified, or revealed revealed through an event evene;;

(ii) Whether prior opportunities opportunities existed existed to identify identifY the problem problem requiring corrective corrective action, and if so, the age and number of those opportunities; (iii) Whether the problem problem was revealed revealed as the result of a licensee self-monitoring licensee self-monitoring effort, such as conducting an audit, a test, a surveillance, troubleshooting; surveillance, a design review, or 'troubleshooting; 9

An "event," as used here, means 9An means (1) an event characterized characterized by an active adverse impact on equipment or or personnel, personnel, readily readily obvious by human human observation observation or instrumentation, instrumentation, or (2)

(2) a radiological radiological impact on personnel or the the environment environment in excess excess of regulatory limits, such as an overexposure, a release of radioactive material above NRC regulatory li,nits, limits, or a loss of radioactive radioactive material.

material. For example, an equipment failure discovered discovered through a spill of liquid, a ofliquid, loud noise, the failure to have a system respond properly, or an annunciator annunciator alarm would be considered considered an event; a system discovered to be inoperable through a document document review would not. Similarly, if a licensee discovered, through quarterly quarterly dosimetry dosimetry readings, that employees employees had been inadequately inadequately monitored for radiation, the issue would normally be considered considered licensee-identified; licensee-identified; however, however, if the same dosimetry readings disclosed an an overexposure, the issue would be considered an event.

23

(iv) For a problem problem revealed through an event, the ease of discovery, and the degree of of licensee licensee initiative in identifying the root cause of the problem and any associated associated violations; (v) For NRC-identified NRC-identified issues"whether issues, whether the licensee licensee would likely have have identified identified the issue in the same time-period time-period if the NRC had not been involved; (vi) For NRC-identified issues, whether the licensee should have identified the issue (and taken action) earlier; and (vii)

(vii) For cases in which the NRC identifies the overall problem requiring requiring corrective action (e.g.,

(e.g., a programmatic programmatic issue), of licensee initiative or lack of initiative in issue), the degree oflicensee identifying the problem or problems requiring corrective identifYing corrective action.

(2) Although some cases may consider all of the above factors, the importance importance of each factor will vary based on the type of case as discussed discussed in the following generalgeneral guidance:

(i) Licensee-Identified. When a problem Licensee-Identified. problem requiring requiring corrective action is licensee-(iLe., identified before the problem has resulted in an event),

identified (i.e., event), the NRC NRC should normally give the licensee credit for actions actions related to identification, regardless of whether prior prior opportunities existed to identifY opportunities identify the problem.

(ii)

(ii) Identified Through an Event. When a problem requiring corrective corrective action is identified identified through an event, the decision decision on whether whether to give the licensee credit for actions related related identification normally should to identification should consider the ease of discovery, whether whether the event occurred as the result of a licensee self-monitoring effort (i.e.,

licensee self-monitoring (i.e., whether the licensee licensee was "looking for the problem"), the degree oflicensee of licensee initiative initiative in identifying identifYing the problem or problems problems requiring corrective corrective action, and whether prior prior opportunities identify the problem.

opportunities existed to identifY Any of these considerations considerations may be overriding overriding if particularly particularly noteworthy noteworthy or particularly particularly egregious. For example,example, if the event occurredoccurred as the result of conducting a surveillance surveillance or similar self-monitoring effort (i.e.,

self-monitoring (i.e., the licensee was looking for the problem), the licensee should normally be given credit credit for identification. Even if the problem was easily discovered discovered (e.g.,

revealed by a large spill ofliquid),

of liquid), the NRC may choose to give credit credit because because noteworthy licensee licensee effort was exerted exerted in ferreting out the root cause and associated associated violations, or simply simply because no prior opportunities opportunities (e.g.,

(e.g., procedural cautions, cautions, post-maintenance post-maintenance testing, quality control

.control failures, readily observable parameter observable parameter trends, or repeated or locked-in annunciator annunciator warnings) existed to identify identifY the problem.

(iii)

(iii) NRC-ldentified. When a problem requiring corrective NRC-Identified. corrective action is NRC-identified, NRC-identified, the decision

.the decision on whether whether to give the licensee licensee credit for actions related related to Identification Identification should should normally normally be basedbased on an additional question: should the licensee have reasonably identified the problem problem (and taken action) earlier?

24 24

In most cases, this reasoning may be based simply on the ease of the NRC inspector's inspector's discovery (e.g., conducting discovery (e.g., conducting a walkdown, observing observing in the control control room, performing performing a confirmatory confirmatory NRC NRC radiation survey, hearing a cavitating cavitating pump, or finding a valve obviously out out of position). In some cases, the licensee's licensee's missed opportunities to identify the problem might might include include a similar previous previous violation, NRC or industry notices, internal audits, or readily observable observable trends.

If the NRC NRC identifies the violation violation but concludes that, under circumstances, the under the circumstances, licensee's licensee's actions related to Identification Identificationwere not unreasonable, unreasonable, the matter would be treated treated as licensee-identified licensee-identified for purposes of assessing the civil penalty. In such cases, the question of of Identification credit shifts to whether the licensee Identification licensee should be penalized penalized for NRC's identification identification ofof the problem.

(iv) Mixed Mixed Identification.

Identification. For "mixed" identification situations (i.e.,

"mixed" identification (i.e., where multiple violations exist, some NRC-identified, NRC-identified, some licensee-identified, licensee-identified, or where where the NRC prompted prompted the licensee to take action that resulted resulted in the identification identification of the violation),

violation), the NRC's evaluation evaluation should should normally determine determine whether the licensee licensee could reasonably have been expected to identify the violation in the NRC's absence. absence. This determination determination should consider, among other things, the timing of the NRC's discovery, the information information available available to the licensee that caused the NRC concern, the specificity specificity of the NRC's concern, the scope of the licensee's licensee's efforts, the level of of licensee resources licensee resources given to the investigation, and whether the NRC's path of analysis had been dismissed or was being pursued in parallel by the licensee.

In some cases, the licensee may have addressed the isolated symptoms of each violation violation (and may have identified the violations), but failed to recognize the common root cause and taken the necessary comprehensive comprehensive action. Where Where this is true, the decision on whether to give licensee licensee credit credit for actions related to Identification Identification should focus on identification problem identification of the problem requiring requiring corrective corrective action action (e.g.,

(e.g., the programmatic programmatic breakdown).

breakdown). As such, dependingdepending on the chronology chronology of the various violations, the earliest earliest of the individual violations violations might be considered missed opportunities for the licensee to have identified identified the larger larger problem.

(v) Missed Missed Opportunities to Identify. Missed opportunities include include prior prior notifications opportunities to identify notifications or missed opportunities identify or prevent violations such as (1) through normal surveillances, surveillances, audits, or quality assurance (QA) activities; (2) through prior notice, i.e.,

activities; (2) i.e., specific NRC or industry notification; or (3) (3) through other reasonable indication of a potential potential problem or violation, such as observations observations of employeesemployees and contractors, and failure to take effective corrective corrective steps. It may include include findings of the NRC, the licensee, or industry made at other facilities operated by the licensee where it is reasonable to expect expect the licensee to take action to identify identify or prevent similar problems at the facility subject to the enforcement enforcement action at issue. In assessing assessing this factor, consideration consideration will be given to, among other things, the opportunities opportunities available available to discover the violation, the ease of discovery, the similarity between between the violation and the notification, the period of time between between when the violation occurred occurred and when the notification notification was issued, the action taken (or planned) planned) by the licensee in response to the 25

notification, and and the the level of management level of management review that the notification received the notification received (or(or should have received).

The evaluation evaluation of missed opportunities should normally missed opportunities normally depend depend on on whether whether the information available information available to the licensee reasonably have should reasonably licensee should have caused action that would have caused action prevented the violation. Missed opportunities prevented applied where opportunities is normally not applied where the licensee licensee appropriately reviewed appropriately reviewed the opportunity application to its activities opportunity for application activities and reasonable action was reasonable action either taken or planned planned to be taken within a reasonable reasonable time.

situations the missed In some situations opportunity is a violation missed opportunity itself. In violation in itself. In these these cases, cases, unless the missed missed opportunity Severity Level opportunity is a Severity Level III III violation missed opportunity violation in itself, the missed opportunity violation violation grouped with the other violations may be grouped single Severity violations into a single Severity Level Level 11III "problem." However, if "problem." However, opportunity is the only violation, then it should not normally the missed opportunity normally be counted counted twice (i.e.,

both as the violation and as a missed counting") unless opportunity--"double counting")

missed opportunity--"double number of unless the number of opportunities missed was particularly opportunities particularly significant.

significant.

timing of the missed opportunity should The timing should also be considered. While While a rigid time-frame is unnecessary, a 2-year period should generally consistency in implementation, generally be considered for consistency as the period reflecting relatively current performance.

reflecting relatively performance.

. (3) When the NRC determines that the licensee should should receive related to receive credit for actions related normally result in either no civil penalty assessment should normally Identification, the civil penalty assessment Identification, penalty or a base civil penalty, based on whether Corrective CorrectiveAction is judgedjudged to be reasonably prompt and comprehensive. When the licensee is not given credit for actions related comprehensive. Identification, the related to Identification, civil penalty assessment normally assessment should normally result in a Notice of Violation with either either a base civil depending on the quality of Corrective escalated by 100 percent, depending penalty or a base civil penalty escalated Corrective licensee's performance Action, because the licensee's performance is clearly not acceptable.

c.

c. Creditfor Credit for Prompt Comprehensive Corrective Prompt and Comprehensive CorrectiveAction The purpose of the Corrective encourage licensees to (1)

CorrectiveAction factor is to encourage (1) take the immediate actions necessary upon discovery of a violation that will restore safety and compliance immediate compliance with the license, regulation(s),

regulation(s), or other requirement(s);

requirement(s); and (2) (2) develop and implement (in a timely manner) the lasting actions that will not only prevent recurrence recurrence of the violation at issue, significance and complexity appropriately comprehensive, given the significance but will be appropriately complexity of the violation, occurrence of violations with similar root causes.

to prevent occurrence Regardless Regardless of other circumstances enforcement history, identification),

circumstances (e.g., past enforcement identification), the licensee's corrective actions should always be evaluated evaluated as part of the civil penalty assessment process. As a reflection of the importance given to this factor, an NRC judgment that the comprehensive will always result in issuing licensee's corrective action has not been prompt and comprehensive at least a base civil penalty.

26

In assessing assessing this factor, consideration consideration will be given to the timeliness of the corrective action (including the promptness action (including developing the schedule for long term corrective promptness in developing corrective action),

action), the adequacy of the licensee's root cause adequacy cause analysis for the violation, and, given the significance significance and complexity of the issue, the comprehensiveness complexity comprehensiveness of the corrective corrective action (i.e., whether the action (i.e., whether action is focused narrowly to the specific violation or broadly to the general general area of concern). Even in in cases cases when the NRC, at the time of the enforcement enforcement conference, identifies additional peripheral or minor corrective corrective action still to be taken, the licensee may be given credit in this area, as long as the licensee's licensee's actions addressed the underlying root cause and are considered sufficient sufficient to prevent recurrence recurrence of the violation and similar violations.

Normally, the judgment of the adequacy of correctivecorrective actions will hinge on whether whether the NRC had to take action to focus the licensee's evaluative corrective process in order to obtain evaluative and corrective comprehensive corrective action. This will normally be judged comprehensive corrective judged at the time of the predecisional predecisional enforcement conference enforcement conference (e.g.,

(e.g., by outlining outlining substantive additional areas where corrective action is needed). Earlier needed). Earlier informal discussions discussions between the licensee and NRC inspectors inspectors or management may result in improved corrective corrective action, but should not normally normally be a basis to deny credit for Corrective Corrective Action. For cases cases in which the licensee does not get credit for actions related to Identification Identificationbecause because the NRC identified the problem, the assessment assessment of the licensee's corrective action should begin from the time when the NRC put the licensee on notice of the problem.

Notwithstanding Notwithstanding eventual comprehensive corrective eventual good comprehensive immediate corrective corrective action, if immediate corrective action action was not taken to restore safetysafety and compliance once compliance once the violation was identified, corrective corrective action would not be considered prompt and comprehensive.

Corrective action action for violations discrimination should violations involving discrimination should normally only be considered comprehensive considered comprehensive if the licensee takes prompt, comprehensive corrective action that comprehensive corrective (I) broader environment for raising safety concerns in the workplac~,

(1) addresses the broader workplace, and (2) provides a remedy for the particular discrimination at issue.

particular discrimination In response response to violations violations of 10 CFR 50.59, corrective corrective action should normally be considered prompt and comprehensive considered comprehensive only if the licensee --

(i) Makes a prompt decision decision on operability; and either (ii) Makes a prompt evaluation under under 10 CFR 50.59 50.59 if the licensee licensee intends to maintain the facility or procedure procedure in the as found condition; condition; or or (iii)

(iii) Promptly initiates corrective corrective action consistent consistent with Criterion XVI of 10 CFR 50, Appendix B, if it intends intends to restore restore the facility or procedure procedure to the FSAR description.

27

d.

d. Exercise Exercise ofDiscretion Discretion provided in Section VII, "Exercise As provided "Exercise of Discretion,"

Discretion," discretion may be exercised by either escalating or mitigating the amount of the civil penalty either penalty determined determined after applying the civil penalty adjustment factors to ensure that the proposed civil penalty reflects all relevant circumstances of circumstances ofthe the particular particular case. However, in no instance will a civil penalty for anyone any one exceed $130,000 violation exceed $130,000 per day.

Orders D. Orders An order is a written written NRC directive directive to modify, --suspend, revoke a license; to cease and suspend, or revoke practice or activity; or to take such other action as may be proper (see desist from a given practice (see 10 CFR 2.202). Orders Orders may also be issued in lieu of, of, or in addition addition to, civil penalties, as appropriate appropriate for Severity Severity Level I, II, or III violations. Orders may be issued as follows:

I.

1. License License Modification orders orders are issued when some change in licensee equipment, equipment, procedures, procedures, personnel, management controls personnel, or management controls is necessary.

2. Suspension Orders Orders may be used:

(a)

(a) To remove a threat to the public health and safety, common defense and security, or the environment; (b) To stop facility construction when, (i)

(i) Fuikher Further work could preclude preclude or significantly identification or correction significantly hinder the identification correction constructed safety-related system or component; of an improperly constructed safety-related component; or or (ii) The licensee's licensee's quality assurance assurance program program implementation implementation is not adequate to provide confidence confidence that construction construction activities are being properly carried carried out; (c) When the licensee licensee has not responded adequately to other enforcement action; (d) When the licensee licensee interferes with the the conduct conduct of an inspection or investigation; or investigation; or (e) For any reason not mentioned mentioned above for which license revocation is legally authorized.

Suspensions may apply to all or part of the licensed licensed activity. Ordinarily, a licensed licensed activity activity is not suspended (nor is a suspension prolonged) prolonged) for failure to comply comply with requirements requirements where such failure is not willful and adequate corrective action has adequate corrective h<l;s been taken.

28

3.

3. Revocation Orders may be used:

Revocation (a)

(a) When a licensee is unable or unwilling to comply with NRC requirements; (b) When a licensee refuses to correct correct a violation; (c)

(c) When licensee does not respond to a Notice of Violation where When licensee where a response was required; (d) When a licensee licensee refuses to pay an applicable fee under the Commission's regulations; or or (e)

(e) For any other reason for which revocation is authorized authorized under section 186 of the Atomic Energy Act (e.g., (e.g., any condition which which would warrant refusal ofa of a license license on an original application).

4. Cease and Desist Orders Orders may be used to stop an unauthorized activity activity that has continued after notification by the NRC that the activity is unauthorized.

continued unauthorized.

5. Orders to non-licensees, including contractors and subcontractors, subcontractors, holders of NRC approvals, e.g., certificates of compliance, e.g., certificates compliance, early site permits, permits, standard standard design certificates, or or applicants for any of them, and to employees employees of any of the foregoing, are used when the NRC has identified deliberate misconduct that may cause a licensee to be in violation of an NRC identified deliberate requirement or where incomplete requirement incomplete or inaccurate inaccurate information is deliberately submitted or where the NRC loses its reasonable assurance that the licensee will meet NRC requirements reasonable assurance requirements with thatthat person involved in licensed activities.

Unless a separate separate response is warranted warranted under 10 CFR 2.201, 2.201, a Notice of Violation need not be issued where an order is based on violations described in the order. The violations described in an order need not be categorized categorized by severity severity level.

Orders are made effective effective immediately, without prior opportunity for hearing, whenever it is determined determined that the public health, interest, or safety so requires, or when the order is Oth~rwise, a prior opportunity responding to a violation involving willfulness. Otherwise, opportunity for a hearing on on the order is afforded. For cases in which which the NRC believes believes a basis could reasonably reasonably exist for not taking the action as proposed, the licenseelicensee will ordinarily be afforded afforded an opportunity opportunity to show why the order should not be issued in the proposed proposed manner by way of a Demand for Information.

(See 10 CFR 2.204) 29

E. RelatedAdministrative Actions Actions In addition addition to NCVs, NOVs, civil to NCVs, civil penalties, penalties, and orders, the NRC also and orders, also uses administrative administrative actions, such asas Notices Notices ofof Deviation, Notices Notices of Nonconformance, ConfIrmatory Action Nonconformance, Confirmatory Letters, Action Letters, Letters of Reprimand, Reprimand, and Demands for Information supplement its enforcement Information to supplement enforcement program.

The NRC expects expects licensees contractors to adhere to any licensees and contractors any obligations obligations and commitments commitments resulting from these actions and will will not hesitate to issue appropriate orders to ensure issue appropriate ensure that these these obligations obligations and commitments are commitments are met.

1. Notices

1. Deviation are written notices describing Notices of Deviation licensee's failure to satisfy describing a licensee's satisfY a commitment where commitment where the the commitment involved has not been made commitment involved made a legally legally binding binding requirement.

requirement.

A Notice of Deviation requests that a licensee of Deviation licensee provide a written written explanation explanation or statement statement corrective steps taken describing corrective describing taken (or planned), achieved, and the date when corrective planned), the results achieved, corrective action will be completed.

2. Notices of Nonconformance Nonconformance are written notices describing contractors'contractors' failures to commitments which have meet commitments requirements by NRC. An have not been made legally binding requirements commitment made example is a commitment example made in a procurement contract with a licensee as required procurement contract required by 10 10 CFR CFR Part 50, Appendix B. Notices of Nonconformances non-licensees provide written Nonconformances request that non-licensees written statements describing explanations or statements explanations describing corrective corrective steps (taken or planned),

planned), the results achieved, the dates when corrective corrective actions will be completed, completed, and measures measures taken to preclude recurrence.

Confirmatory Action Letters are letters confIrming

3. Confirmatory

3. confirming a licensee's or contractor's significant concerns agreement to take certain actions to remove signifIcant agreement concerns about health and safety, safeguards, or the environment.

safeguards,

4. Letters of Reprimand are letters addressed to individuals subject subject to Commission identifYing a signifIcant jurisdiction identifying significant defIciency deficiency in in their performance performance of licensed activities.

Information are demands

5. Demands for Information

5. demands for information from licensees or other enabling the NRC to determine whether an order or other enforcement persons for the purpose of enabling enforcement action should be issued.

VII. EXERCISE OF DISCRETION Notwithstanding the normal guidance contained in this policy, as provided in Section Section III, "Responsibilities," the NRC may choose to exercise discretion and either escalate or mitigate "Responsibilities,"

enforcement sanctions within the Commission's statutory authority to ensure that the resulting particular circumstances of the particular enforcement action takes into consideration all of the relevant circumstances case.

30 30

A. Escalation A. Escalationof Enforcement Enforcement Sanctions Sanctions The NRC considers violations categorized categorized at Severity Level I, II, I1, or III to be of significant regulatory regulatory concern. The NRC also considers considers violations associated with findings that the Reactor Reactor Oversight Process's Oversight Process's Significance Detennination Process evaluates as having low to moderate, or Significance Determination greater greater safety significance (i.e.,

(i.e., white, yellow, or red) to be of significant regulatory concern. If If the application application of the normal guidance in this policy does not result in an appropriate appropriate sanction, with the approval approval of the Deputy Executive Director and consultation with the EDO and Commission, as warranted, the NRC NRC may apply its full enforcement enforcement authority where the action action is warranted. NRC action may include (1) escalating escalating civil penalties; (2)(2) issuing appropriate appropriate orders; and (3)

(3) assessing civil penalties penalties for continuing continuing violations on a per day basis, up to the statutory of $130,000 per violation, per day.

limit of$130,000

1. Civil Penalties Civil Penalties Notwithstanding Notwithstanding the outcome of the normal civil penalty assessment process addressed penalty assessment addressed in Section VI.C, the NRC may exercise discretion discretion by either proposing a civil penalty where application of the factors would otherwise otherwise result in zero penalty or by escalating the amount of of the resulting resulting civil penalty (i.e.,

(i.e., base or twice the base civil penalty) to ensure ensure that the proposed civil penalty reflects the significance significance of the circumstances.

circumstances. The Commission will be notified if notified if the deviation deviation in the amount of the civil penalty proposed under this discretion from the amount of of the civil penalty assessed under the normal process is more than two times the base civil penalty penalty shown in Tables lA 1A and IB.

lB. Examples when this discretion should be considered include, but are not limited to the following:

(a) Problems categorized categorized at Severity Level lor I or II; 1I; (b) Overexposures, or releases of radiological Overexposures, radiological material material in excess of NRC requirements; requirements; (c)

(c) Situations involving particularly poor licenseelicensee performance, performance, or involving involving willfulness; willfulness; (d) Situations when the licensee's licensee's previous enforcement history has been previous enforcement been particularly poor, or when the current current violation violation is directly repetitive of an earlier violation; (e)

(e) Situations when the violation results in a substantial substantial increase in risk, including cases in which the duration of the violation has contributed contributed to the substantial increase; increase; (f) Situations when the licensee licensee made a conscious decision to be in noncompliance noncompliance in order to obtain an economic economic benefit; 31 31

(g) Cases involving the loss, abandonment, or improper transfer transfer or disposal of a source or device. Notwithstanding the outcome sealed source outcome of the normal civil penalty assessment assessment cases normally should result in a civil penalty of at least the base amount; or process, these cases or (h) Levei II or III violations associated Severity Level Severity associated with departures departures from the Final Safety Safety Analysis Report identified after March Analysis March 30,2000, 30, 2000, for risk-significant items as defined by the maintenance rule program and March licensee's maintenance March 30, 2001, for all other issues. Such a violation 30,2001, violation or problem would consider consider the number and nature nature of the violations, the severity ofthe violations, of the whether the violations were continuing, and who identified whether identified the violations (and if the licenseelicensee identified the violation, whether whether exercise VII.B.3 enforcement exercise of Section VILB.3 enforcement discretion is warranted.)

warranted. )

Orders

2. Orders The N~C necessary or desirable, issues orders in conjunction with or in lieu NRC may, where necessary of civil penalties penalties to achieve or formalize corrective corrective actions and to deter further recurrence recurrence of of serious violations.

Civil Penalties

3. Daily Civil Penalties In order to recognize significance for those cases where recognize the added significance where a very strong message message warranted for a significant violation that continues for more than one day, the NRC may is warranted violation and attendant civil penalty up to the statutory exercise discretion and assess a separate violation violation continues. The NRC may exercise

$130,000 for each day the violation limit of $130,000 exercise this discretion if a licensee was aware of or clearly should should have been aware aware of a violation, or if the licensee had an an identify and correct the violation but failed to do so.

opportunity to identifY B. Mitigation B. Enforcement Sanctions Mitigationof Enforcement Sanctions The NRC may exercise discretion and refrain refrain from issuing a civil penalty and/or a Notice considering the general principles of this statement of policy and the Violation after considering of Violation 0

surrounding circumstances.1O surrounding circumstances." The approval approval of the Director, Office of Enforcement, Enforcement, in consultation consultation with the Deputy Executive Director, as warranted, is required for exercising exercising discretion of the type described in Sections VILB.2 through VII.B.6.

Sections VII.B.2 VILB.6. The circumstances circumstances under which mitigation discretion discretion should be considered considered include, but are not limited to the following:

[Reserved]

1. [Reserved]

The mitigation 0 The 10 mitigation discretion described in discretion described in Sections Sections VII.B.2 V11.B.2 -- V1I.B.6 does not VII.B.6 does normally apply not normally to violations apply to violations associated with issues evaluated by the SDP. The Reactor Oversight Oversight Process will use the Agency Action Matrix to determine the agency agency response to performance performance issues. The Agency Action Matrix Matrix has provisions to consider provisions consider circumstances that were previously addressed through enforcement extenuating circumstances enforcement mitigation.

32 32

2. Violations Identified

2. Violations Identified During Extended Shutdowns During Extended Shutdowns or Work Stoppages Stoppages The NRC may refrain refrain from issuing a Notice Notice of Violation or a proposed proposed civil penalty for a Severity Severity Level Level II, III, or IV violation that is identified identified after (i) the NRC has taken significant significant enforcement action based enforcement based upon a major safety event contributing contributing to an extended extended shutdown of an an operating operating reactor reactor or a material material licensee (or a work stoppage at a construction construction site), or (ii) the licensee enters an extended shutdown or work stoppage stoppage related to generally poor performance performance over a long period oftime, provided that the violation is documented in an inspection of time, provided inspection report (or (or inspection records for some material cases) and that it meets all of the following criteria:

inspection (a) It was either licensee-identified licensee-identified as a result ofaof a comprehensive comprehensive program program for for problem identification correction that was developed identification and correction developed in response response to the shutdown or or identified as a result of an employee allegation allegation to the licensee; (If licensee; (If the NRC identifies identifies the violation violation and all of the other criteria ofthe criteria are met, the NRC should determine whether enforcement enforcement action is necessary to achieve remedial remedial action, or if discretion may still be appropriate.)

appropriate.)

(b) It is based upon activities oftheof the licensee prior to the events leading to the shutdown; (c)

(c) categorized at Severity Level I; It would not be categorized (d) It was not willful; and (e)

(e) The licensee's licensee's decision to restart the plant requires NRC concurrence.

concurrence.

3. Violations

3. Violations Involving Old DesignDesign Issues Issues The NRC may refrain from proposing a civil penalty penalty for a Severity Severity Level II IL or III violation involving a past problem, such as in engineering, engineering, design, or installation, ifthe if the violation is documented in an inspection report (or inspectioninspection records records for some material cases) that includes includes a description description of the corrective corrective action and that it meets all of the following criteria:

(a) It was a licensee-identified licensee-identified as a result of its voluntary voluntary initiative; (b) It was or will be corrected, including immediate immediate corrective action and long term comprehensive comprehensive corrective action to prevent recurrence, within a reasonable time following identification (this action should involve expanding identification expanding the initiative, as necessary, to identify other failures caused by similar root causes);

causes); and (c) It was not likely to be identified (after the violation occurred) by routine licensee routine licensee efforts efforts such as normal surveillance surveillance or quality assurance (QA) activities.

33

Violation for a Severity refrain from issuing a Notice of Violation In addition, the NRC may refrain Severity 1II, or IV violation that meets the above criteria Level II, III, provided the violation was caused by criteria provided conduct that is not reasonably linked to present performance (normally, violations that are at least present performance 3 years old or violations occurring during plant construction) and there had not been prior notice violations occurring identified the violation earlier. This exercise of so that the licensee should have reasonably identified of licensees initiating efforts to identify and correct discretion is to place a premium on licensees correct subtle routine efforts before degraded safety systems are violations that are not likely to be identified by routine called upon to work.

Section VII.B.3 discretion would not normally applied to departures from the FSAR if:

normally be applied if:

(a)

(a) The NRC identifies the violation, unless it was likely in the NRC staff's staffs view that the licensee would have identified the violation in light of the defined scope, thoroughness, thoroughness, and completion of the schedule of the licensee's initiative provided the schedule provides for completion schedule risk-significant items as defined by the licensee's licensee's initiative by March 30, 2000, for risk-significant licensee's maintenance rule program and by March 30, maintenance 2001, for all other issues; 30,2001, (b) The licensee identifies the violation violation as a result of an event other surveillance or other event or surveillance identifies the FSAR issue; - "

required testing where corrective action required corrective where required action identifies the FSAR issue; (c)

(c) The licensee identifies the violation but had prior opportunities to do so (was aware of the departure departure from the FSAR) and failed to correct it earlier; (d) associated with the violation; There is willfulness associated (e)

(e) The licensee fails to make a report required required by the identification identification of the departure from the FSAR; or or (f) comprehensive corrective action or fails to The licensee either fails to take comprehensive appropriately expand the corrective appropriately corrective action should be broad corrective action program. The corrective broad with a defined scope and schedule.

Violations Identified

4. Violations Identified Due to Previous Enforcement Action Previous Enforcement The NRC may refrain from issuing a Notice of Violation or a proposed civil penalty for a Severity Level II, III, or IV violation that is identified after the NRC has taken enforcement enforcement action, if the violation violation is documented inspection records for some documented in an inspection report (or inspection cases) that includes a description material cases) corrective action and that it meets all of the description of the corrective following criteria:

(a) It was licensee-identified corrective action for the previous licensee-identified as part of the corrective enforcement enforcement action; 34 34

(b) It has the same or similar root cause as the violation violation for which enforcement enforcement action action was issued; (c)

(c) It does not substantially change the safety significance significance or the character character of the regulatory concern arising out of the initial violation; and (d) It was or will be corrected, corrected, including ~ediate immediate corrective corrective action and long term comprehensive corrective comprehensive corrective action to prevent prevent recurrence, recurrence, within a reasonable reasonable time following identification.

(e)

(e) It would not be categorized categorized at Severity Levell; Level I;

5. Violations Violations Involving Certain CertainDiscrimination Discrimination Issues Issues Enforcement discretion may be exercised Enforcement exercised for discrimination discrimination cases when a licensee who, without the need for government government intervention, identifies an issue of discrimination discrimination and takes prompt, comprehensive, comprehensive, and effective corrective action to address both the particular situation situation and the overall work environment for raising safety concerns. Similarly, enforcement enforcement may not be warranted where a complaint is filed with the Department warranted Department of Labor (DOL) under Section 211 of of the Energy Reorganization Reorganization Act of 1974,1974, as amended, but the licensee settles the matter before the DOL makes an initial finding of discrimination discrimination and addresses the overall overall work environment.

environment.

Alternatively, if a finding of discrimination is made, the licensee may choose to settle the case before the evidentiary evidentiary hearing begins. In such cases, the NRC may exercise exercise its discretion not to take enforcement enforcement action when the licensee has addressed addressed the overall work environment environment for raising safety concerns concerns and has publicized that a complaint complaint of discrimination discrimination for engaging in in protected activity was made to the DOL, that the matter was settled to the satisfaction satisfaction of the employee employee (the terms of the specific settlement settlement agreement need not be posted), and that, if the DOL Area Office found discrimination, the licensee has taken action to positively positively reemphasize reemphasize that discrimination will not be tolerated. Similarly, the NRC may refrain refrain from taking enforcement enforcement action if a licensee settles a matter promptly after a person comes to the NRC without going to the DOL. Such discretiondiscretion would normally not be exercised exercised in cases in which which the licensee does not appropriately appropriately address the overall overall work environment (e.g., by using training, environment (e.g.,

postings, po stings, revised policies policies or procedures, procedures, any necessary necessary disciplinary etc., to communicate disciplinary action, etc., communicate its policy against discrimination) or in cases that involve: allegations of discrimination as a result of providing information directly to the NRC, allegations of discrimination discrimination caused by a manager manager above first-line supervisor (consistent (consistent with current Enforcement Enforcement Policy classification Severity classification of Severity I or II violations), allegations Level lor discrimination where a history offmdings allegations of discrimination of findings ofof discrimination (by the DOL or the NRC) or settlements suggests a programmatic programmatic rather than an isolated discrimination discrimination problem, or allegations allegations of discrimination which appear particularly blatant or egregious.

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6. Violations

6. Circumstances Violations Involving Special Circumstances Notwithstanding Notwithstanding the outcome enforcement process normal enforcement outcome of the normal process addressed addressed in Section VI.B Section Vl.B oror the normal normal civil penalty penalty assessment process addressed assessment process addressed inin Section Vl.C, the NRC Section VI.C, may reduce refrain from issuing a civil penalty or a Notice reduce or refrain Notice ofof Violation Severity Level II, Violation for a Severity II,

III, III, or IV violation IV violation based on the merits of the case case after considering considering the guidance guidance in this statement statement of policy and of and such factors as the age of the such factors the violation, the significance significance of the violation, violation, the clarity clarity of the requirement, ofthe requirement, the appropriateness of the requirement, the appropriateness sustained performance of requirement, the overall sustained performance of the licensee has been particularly the licensee particularly good, and and other relevant circumstances, including any relevant circumstances, any that may may have changed since the violation. This discretion is have changed is expected expected to be exercised only be exercised only where application of the application the normal guidance in normal guidance unwarranted. In addition, the in the policy is unwarranted. the NRC may refrain from issuing refrain issuing enforcement enforcement action for violations resulting from matters matters not not within within a. a.

licensee's licensee's control, such as equipment were not avoidable by reasonable equipment failures that were reasonable licensee licensee quality quality assurance measures management controls. Generally, however, licensees measures or management licensees are held held responsible for the acts of their employees responsible contractors. Accordingly, this policy employees and contractors. policy should not not be construed to excuse excuse personnel personnel or contractor contractor errors.

C. Notice of Enforcement Enforcement Discretion Discretion for for Power Power Reactors Reactors and Diffusion Plants Gaseous Diffusion and Gaseous Plants On occasion, circumstances may arise where a power circumstances may compliance with a power reactor's compliance Technical (TS) Limiting Condition Specification (TS)

Technical Specification Operation or with other license conditions Condition for Operation unnecessary plant transient would involve an unnecessary performance of testing, inspection, or system transient or performance system realignment that is inappropriate with the specific realignment specific plant conditions, unnecessary delays in plant conditions, or unnecessary corresponding health and safety benefit. Similarly, for a gaseous diffusion plant startup without a corresponding (GDP), circumstances compliance with a Technical Safety Requirement (TSR) circumstances may arise where compliance specification or other certificate condition would unnecessarily or technical specification unnecessarily call for a total plant shutdown or, notwithstanding that a safety, safeguards, or security feature was degraded or or inoperable, compliance inoperable, unnecessarily place the plant in a transient or condition where compliance would unnecessarily those features could be required.

circumstances, the NRC staff may choose not to enforce the applicable TS, TSR, In these circumstances, enforcement discretion, designated as a Notice of or other license or certificate condition. This enforcement of Enforcement Enforcement Discretion (NOED), (NOED), will only be exercised if the NRC staffis staff is clearly satisfied that the action is consistent with protecting the public health and safety. The NRC staff may also enforcement discretion in cases involving severe weather or other natural phenomena, grant enforcement phenomena, based upon balancing balancing the public health and safety or common defense and security of not operating against the potential associated with continued potential radiological or other hazards associated continued operation, unacceptably by exercising this discretion.

and a determination that safety will not be impacted unacceptably The Commission is to be informed expeditiously following the granting ofa informed expeditiously of a NOED in these situations. A licensee or certificate holder seeking the issuance of a NOED must proyide provide a circumstances where good cause is shown, oral justification written justification, or in circumstances justification followed as soon as possible by written justification, that documents the safety basis for the request and 36 36

provides whatever other information necessary necessary for the NRC staff to make a decision on whether whether to issue a NOED.

For power reactors, the appropriate appropriate Regional Administrator, or his or her designee, may issue a NOED after consultation consultation with the Director, Office of Nuclear Reactor Regulation, or or his or her designee, to determine determine the appropriateness of granting a NOED where (1) the noncompliance is temporary noncompliance temporary and nonrecurring nonrecurring when an amendment amendment is not practical practical or or (2) if the expected noncompliance (2) noncompliance will occur during the brief period of time it requires requires the NRC staff to process process an emergency emergency or exigent amendment under the provisions exigent license amendment provisions of 10 CFR 50.91 (a)(5)

(a)(5) or (6).

(6). For gaseous diffusion plants, the appropriate appropriate Regional Administrator, Administrator, or his or her designee, may issue and document document a NOED where the noncompliance noncompliance is temporary temporary and and nonrecurring nonrecurring and when an amendment amendment is not practical. The Director, Office of Nuclear Materials Safety and Safeguards, Safeguards, or his or her designee, may issue a NOED ifthe if the expected expected noncompliance noncompliance will occur occur during the brief period of time it requires the NRC staff to process a certificate certificate amendment under 10 CFR 76.45. The person exercising amendment exercising enforcement enforcement discretion discretion will document the decision.

For an operating operating reactor, this exercise exercise of enforcement enforcement discretion discretion is intended to minimize minimize the potential potential safety consequences of unnecessary consequences unnecessary plant transients with accompanying the accompanying operational risks and impacts or to eliminate operational eliminate testing, inspection, or system realignment which is inappropriate for the particular inappropriate particular plant conditions. For plants in a shutdown condition, exercising exercising enforcement discretion is intended to reduce shutdown enforcement inspection shutdown risk by, again, avoiding testing, inspection or system realignment which is inappropriate inappropriate for the particular particular plant conditions, in that, it does not provide provide a safety benefit benefit or may, in fact, be detrimental detrimental to safety in the particular particular plant condition. Exercising Exercising enforcement discretion enforcement discretion for plants attempting to startup is less likely than exercising it for an operating plant, as simply delaying startup does not usually leave the plant in in a condition condition in which it could experience experience undesirable undesirable transients. In such cases, the Commission Commission would expect that discretion would be exercised exercised with respect to equipment or systems only when when it has at least concluded concluded that, notwithstanding notwithstanding the conditions of the license:

license: (1) the equipment or system does not perform perform a safety function in the mode in which operation operation is to occur; (2) the occur; (2) performed by the equipment safety function performed equipment or system is of only marginal safety benefit, benefit, provided remaining remaining in the current mode increases the likelihood of an unnecessary plant unnecessary transient; or (3)

(3) the TS or other license condition requires a test, inspection, or system realignment that is inappropriate inappropriate for the particular particular plant conditions, in that it does not provide a safety benefit, or may, in fact, be detrimental to safety in the particular plant condition.

For GDPs, the exercise of enforcement enforcement discretion would be used where compliance compliance with a certificate certificate condition would involve an unnecessary unnecessary plant shutdown shutdown or, notwithstanding notwithstanding that a safety, safeguards, safeguards, or security feature was degraded.or degraded or inoperable, compliance compliance would unnecessarily place the plant in a transient or condition where unnecessarily where those features could could be required.

Such regulatory flexibility is needed needed because because a total plant shutdown shutdown is not necessarily necessarily the best response to a plant condition. GDPs are designed to operate continuously continuously and have never been been shut down. Although portions can be shut down for maintenance,maintenance, the NRC staff has been 37 37

informed informed by by the the certificate certificate holder holder that restart restart from from aa total plant plant shutdown shutdown maymay not not be practical practical and the the staff staff agrees agrees that the design design of a GDP does not make make restart restart practical.

practical. Hence, Hence, the decision decision to place place either either GDP in in plant-wide plant-wide shutdown shutdown condition condition would would be made made only after after determining determining that that there there is is inadequate inadequate safety, safeguards, safeguards, or or security security and and considering considering thethe total total impact impact ofof the shutdown shutdown on safety, the environment, environment, safeguards, safeguards, and security. A NOED would would not be used for for noncompliances noncompliances with other than certificate requirements, other than certificate requirements, or or for situations where situations where the certificate certificate holder cannot demonstrate adequate cannot demonstrate safeguards, or adequate safety, safeguards, or security.

The decision to exercise exercise enforcement enforcement discretion discretion does not not change change the fact that a violation violation will occur nor nor does it imply that enforcement enforcement discretion is is being exercised exercised for any any violation that may have led to the violation violation at issue. In each case where the NRC staff has chosen chosen to issue a NOED, enforcement NOED, enforcement action will normally normally be taken for the rootroot causes, causes, to the extent violations were involved, that led to the noncompliance noncompliance for which enforcement enforcement discretion discretion was used. The enforcement action enforcement action is intended intended to emphasize licensees and certificate emphasize that licensees certificate holders should should not rely on Qn the NRC's authority to exercise enforcement discretion exercise enforcement discretion as a routine routine substitute substitute for compliance compliance or for requesting requesting a license or certificate certificate amendment.

amendment.

38 38

Finally, it is expected expected that the NRC staff will exercise enforcement enforcement discretion in this area area infrequently. Although a plant must shut down, refueling activities may be suspended, suspended, or plant startup startup may be delayed, absent the exercise of enforcement enforcement discretion, the NRC staff is under no obligation obligation to take such a step merely because it has been requested. The decision to forego enforcement enforcement is discretionary. When enforcement enforcement discretion is to be exercised, exercised, it is to be exercised exercised only if the NRC staff is clearly clearly satisfied that the action is warranted warranted from a health and and safety perspective.

perspective.

VIII. ENFORCEMENT ENFORCEMENT ACTIONS INVOLVING INDIVIDUALS ACTIONS INVOLVING INDIVIDUALS Enforcement actions involving individuals, including Enforcement significant including licensed operators, are significant personnel personnel actions, which which will be closely controlled and judiciously judiciously applied. An enforcement enforcement action involving an individual will normally nOfI11ally be taken only when the NRC is satisfied that the individual fully understood, or should have understood, his or her responsibility; knew, or should have have known, the required actions; and knowingly, or with careless disregard (i.e., with more than disregard (i.e.,

mere mere negligence) negligence) failed to take required required actions which have actual or potentialpotential safety significance. Most transgressions of individuals individuals at the level of Severity Severity Level III or IV violations will be handled handled by citing only the facility licensee.

More serious violations, including including those involving the integrity integrity of an individual (e.g.,

lying to the NRC) concerning concerning matters within the scope of the individual's individual's responsibilities, responsibilities, will be considered for enforcement enforcement action against the individual as well well as against the facility licensee.

However, action against the individual will not be taken if the improper action by the individual was caused caused by management management failures. The following examples of situations illustrate illustrate this concept:

• • Inadvertent individual mistakes resulting from inadequate Inadvertent inadequate training or guidance guidance provided provided by the facility licensee.

licensee.

• 0 Inadvertently missing an insignificant procedural Inadvertently procedural requirement requirement when the action action is routine, fairly uncomplicated, uncomplicated, and there is no unusual circumstance circumstance indicating indicating that the procedures procedures should be referred referred to and followed step-by-step.

• 0 Compliance with an express direction Compliance direction of management, such as the Shift Supervisor or Plant Manager, resulted in a violation unless the individual did not express his or her concern concern or objection objection to the direction.

• 0 Individual Individual error directly resulting from following the technical technical advice of an expert unless the advise was clearly unreasonable and the licensed clearly unreasonable licensed individual individual should have recognized it as such.

• Violations resulting from inadequate inadequate procedures procedures unless the individual individual used a procedure knowing it was faulty and had not attempted faulty procedure attempted to get the procedure corrected.

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Listed below are examples of situations situations which could result in enforcement enforcement actions involving individuals, licensed or unlicensed. If the actions described in these examples are taken by a licensed operator operator or taken deliberately deliberately by an unlicensed individual, enforcement enforcement action action may be taken may be taken directly against the individual. However, violations involving willful conduct not amounting amounting to deliberate deliberate action by an unlicensed individual individual in these situations may result in enforcement enforcement action against a licensee that may impact an individual. The situations include, but are not limited to, violations that involve:

• " Willfully causing a licensee to be in violation of NRC requirements.

• " Willfully taking action that would have caused a licensee to be in violation of NRC requirements but requirements but the the action did not do so becausebecause it was detected corrective action was taken.

detected and corrective

" Recognizing

• Recognizing a violation of procedural requirements requirements and willfully not taking corrective action.

• " Willfully defeating defeating alarms which have safety significance.

• Unauthorized

• Unauthorized abandoning abandoning of reactor controls.

• " Dereliction Dereliction of duty.

• Falsifying FalsifYing records required by NRC regulations or by the facility license.

• • Willfully providing, or causing a licensee to provide, an NRC inspector investigator inspector or investigator inaccurate or incomplete with inaccurate incomplete information information on a matter material to the NRC.

• 0 Willfully withholding withholding safety safety significant significant information information rather than making such information information known to appropriate supervisory or technical personnel personnel in the licensee's organization.

• 9 Submitting false information information and as a result gaining unescorted access to a nuclear nuclear power plant.

• • Willfully providing providing false data to a licensee by a contractor contractor or other person who provides test provides test or other services, when the data affects the licensee's compliance with 10 CFR licensee's compliance Part 50, Appendix B, or other regulatoryregulatory requirement.

• • Willfully providing providing false certification certification that components meet the requirements of their intended use, such as ASME Code.

• • Willfully supplying, by contractors contractors of equipment for transportation of radioactive radioactive material, casks that do not comply with their certificatescertificates of compliance.

compliance.

• • Willfully performing performing unauthorized unauthorized bypassing of required required reactor or other facility safety systems.

• • Willfully taking actions that violate Technical Specification Specification Limiting Conditions for Operation or other license conditions (enforcement action for a willful violation conditions (enforcement violation will not be taken if taken if that that violation violation is the result of action taken following the NRC's decision to forego enforcement of the Technical enforcement Technical Specification Specification or other license condition or if the operator meets the requirements requirements of 10 CFR 50.54 (x),

(x), (i.e., unless the operator operator acted unreasonably considering considering all circumstances surrounding the relevant circumstances surrounding the emergency.)

emergency.)

Normally, some enforcement enforcement action is taken against a licensee for violations caused caused by by significant acts significant acts of wrongdoing by its employees, contractors, or contractors' employees. In deciding whether deciding whether to issue issue an enforcement enforcement action to an unlicensed person as well as to the 40

licensee, licensee, the NRCNRC recognizes that judgments will have to be made on a case that judgments case by by case basis. In In making these making these decisions, decisions, the NRCNRC will consider consider factors such as the following:following:

I1.

1. The level individual within the organization.

level of the individual organization.

2. The individual's individual's training training and experience experience as well as knowledge knowledge of the potential potential consequences consequences of the wrongdoing.

wrongdoing.

3. The safety safety consequences consequences of the misconduct.

misconduct.

4. The benefit benefit to the wrongdoer, wrongdoer, e.g., e.g., personal personal or corporate corporate gain.

5.

5. The degree supervision of the individual, degree of supervision *individual,i.e.,

i.e., how closely is the individual individual monitored monitored or audited, likelihood of detection audited, and the likelihood detection (such as a radiographer radiographer working working independently in independently in the field as contrasted contrasted with a team activity at a power plant). plant).

6. The employer's employer's response, response, e.g., disciplinary action taken.

e.g., disciplinary

7. The attitude attitude of the wrongdoer, e.g., wrongdoing, acceptance e.g., admission of wrongdoing, acceptance of of responsibility.

responsibility.

8.

8. The degree degree of management management responsibility culpability.

responsibility or culpability.

9. Who identified misconduct.

identified the misconduct.

Any proposed enforcement action involving individuals proposed enforcement individuals must be issued with with the concurrence concurrence of the Deputy Executive Executive Director. The particular particular sanction sanction to be used should should bebe determined determined on a case-by-case basis."'

case-by-case basis. I I Notices Notices of Violation Violation and Orders are examples examples of of enforcement enforcement actions that may be appropriate against individuals.

appropriate against individuals. The administrative administrative action action of a Letter Reprimand may also be considered.

Letter of Reprimand considered. In addition, the NRC may issue Demands for Information to gathergather information to enable it to determine determine whether an order or other other enforcement action should be issued.

enforcement Orders to NRC-licensed reactor operators NRC-licensed reactor operators may involve involve suspension suspension for a specified period, specified period, modification, revocation of their individual modification, or revocation 'individual licenses. Orders to unlicensed unlicensed individuals individuals might might include provisions provisions that would:

1 "IIExcept Except for individuals subject subject to civil penalties penalties under under section section 206 of the Energy Energy Reorganization Reorganization Act Act of 1974, 1974, as

amended, amended, the NRC will not normally impose impose a civil civil penalty against against an individual. However, section individual. However, section 234 of the Atomic Atomic Energy Energy Act (AEA)

(AEA) gives gives the Commission authority to impose penalties on "any person."

impose civil penalties person.' "Person"

'Person" is*

is broadly broadly defined in Section I Is of the ABA Section lIs AEA to include include individuals, a variety of organizations, organizations, and any representatives representatives or agents. This gives the Commission authority authority to impose impose civil civil penalties penalties on employees employees oflicensees of licensees or on separate separate entities entities when when a violation violation of a requirement requirement directly directly imposed imposed on them is committed.

committed.

41

• 0 Prohibit involvement in NRC licensed activities for a specified period of time (normally the period of suspension suspension would not exceed exceed 5 years) or until certain certain conditions are satisfied, e.g.,

e.g., completing completing specified specified training training or meeting certain certain qualifications.

• • Require notification to the NRC before resuming Require notification resuming work in licensed activities.

• • Require the person to tell a prospective prospective employer or customercustomer engaged engaged in licensed licensed activities that the person activities person has been subject to an NRC order.

In the case of a licensed operator's operator's failure to meet applicable applicable fitness- for-duty requirements fitness-for-duty requirements (10 CFR 55.53(j)),

(10 55.530)), the NRC may issue a Notice of Violation or a civil penalty to the Part 55

licensee, licensee, or an order to suspend, modify, or revoke revoke the Part 55 license. These actions may be taken the first time a licensed operator operator fails a drug or alcohol test, that is, receivesreceives a confirmed positive test that exceeds the cutoff levels of 10 CFR Part 26 or the facility licensee's cutoff cutoff levels, if lower.

iflower. However, normally normally only a Notice of Violation Violation will be issued for the first confirmed confirmed positive positive test in the absence aggravating circumstances absence of aggravating circumstances such as errors in the performance of performance licensedduties or evidence oflicensedduties evidence of prolonged use. In addition, the NRC intends to issue an order to suspend the Part 55 license for up to 3 3 years the second time a licensed operator operator exceeds exceeds those cutoff cutoff levels. In the event there are less than 3 3 years remaining remaining in the term of the individual's license, the NRC NRC may consider consider not renewing renewing the individual's license or not issuing a new license after the three year period is completed. The NRC intends to issue an order to revoke the Part 55 license the third time a licensed operator operator exceeds exceeds those cutoff levels. A licensed operator licensed operator or applicant who refuses to participate participate in the drug and 'alcohol alcohol testing programs established by the facility licensee established licensee or who is involved in the sale, use, or possession possession of an illegal drug is also subject to license license suspension, revocation, or denial.

In addition, the NRC may take enforcementenforcement action against a licensee that may impact impact an an individual, wherewhere the conduct of the individual places in question question the NRC's reasonable reasonable assurance assurance that licensed activities will be properly properly conducted. The NRC may take enforcement enforcement action action for reasons reasons that would warrant refusal to issue a license on an original application.

Accordingly, appropriate enforcement enforcement actions may be taken regarding matters that raise issues of of integrity, competence, competence, fitness-for-duty, or other matters that may not necessarily be a violation of of specific Commission Commission requirements.

In the case of an unlicensed unlicensed person, whether a firm or an individual, an order modifying modifying the facility license may be issued to require require (1) the removal of the ofthe person from all licensed licensed activities activities for a specified period of time or indefmitely, oftime indefinitely, (2)

(2) prior notice to the NRC before using the person person in licensed activities, or (3) (3) the licensee to provide notice ofthe of the issuance issuance of such an order to other persons involved in licensed licensed activities making making reference reference inquiries. In addition, orders to employers employers might require retraining, additional oversight, additional oversight, or independent verification of independent verification of activities performed performed by the person, if the person is to be involved in licensed activities.

42

IX. INACCURATE lNACCURA TE AND AND INCOMPLETE INFORMATION INFORMATION A violation of the regulations involving the submittal of incomplete inaccurate incomplete and/or inaccurate information, whether whether or not considered considered a material false statement, statement, can result in the full range of of enforcement sanctions. The labeling enforcement labeling of a communication communication failure as a material material false statement will be made on a case-by-case case-by-case basis and will be reserved for egregious violations. Violations inaccurate or incomplete involving inaccurate incomplete information or the failure to provide significant information significant information identified by a licensee normally normally will be categorized categorized based on the guidance guidance herein, in Section IV, "Significance of Violations,"

"Significance Violations," and in Supplement VII.

The Commission recognizes recognizes that oral information information may in some situations be inherently less reliable than written submittals becausebecause of the absence of an opportunity for reflection and management management review. However,However, the Commission must be able to rely on oral communications communications from licensee officials officials concerning significant significant information. Therefore, Therefore, in determining whetherwhether to enforcement action take enforcement consideration may be given to factors such as action for an oral statement, consideration knowledge that the communicator should have had, regarding (1) the degree of knowledge regarding the matter, in view of his or her position, experience; (2) position, training, and experience; (2) the opportunity opportunity and time available available prior communication to assure the accuracy or completeness to the communication completeness of the information; (3) (3) the degree of intent or negligence, negligence, if any, involved; (4)

(4) the formality of the communication; (5) communication; (5) the reasonableness of NRC reliance on the information; reasonableness information; (6) the importance of the information information which which was wrong or not provided; provided; and (7)

(7) the reasonableness reasonableness of the explanation explanation for not providing complete and accurate information.

Absent at least careless careless disregard, disregard, an incomplete or inaccurate inaccurate unsworn oral statement statement normally will not be subject to enforcement action unless it involves significant significant information information provided provided by a licensee official. However, enforcement enforcement action may be taken for an unintentionally unintentionally incomplete incomplete or inaccurate inaccurate oral statement statement provided to the NRC by a licensee official or others on behalf of a licensee, licensee, if a record record was made of the oral information and provided provided to the licensee thereby thereby permitting an opportunity opportunity to correct correct the oral information, information, such as if a transcript ofthe of the communication communication or meeting meeting summary containing the error was made available subsequently corrected to the licensee and was not subsequently corrected in a timely manner.

inaccurate or incomplete When a licensee has corrected inaccurate incomplete information, the decision to issue a Notice of Violation for the initial inaccurate inaccurate or incomplete information nonpally information normally will be dependent on the circumstances, circumstances, including the ease ease of detection of the error, the timeliness of the correction, whether the NRC or the licensee licensee identified the problem problem with the communication, and whether whether the NRC relied on the information prior to the correction. Generally, if the matter was promptly promptly identified identified and corrected corrected by the licensee prior prior to reliance by the NRC, or before the NRC NRC raised a question about the information, no enforcement enforcement action will be taken for the initial inaccurate inaccurate or incomplete incomplete information. On the other hand, if the misinformation misinformation is identified after after the NRC relies on it, or after some question question is raised regarding regarding the accuracy of the information, then some enforcement enforcement action normally will be taken even if it is in fact corrected. corrected. However, if if the initial initial submittal was accurate accurate when made but later turns out to be erroneous erroneous because of newly 43

discovered information or advance advance in technology, a citation citation normally would not be appropriate appropriate if,if, when the new information became available became available or the advancement in technology advancement technology was made, the initial submittal was corrected.

The failure to correct correct inaccurate or incomplete incomplete information information which the licensee does not identify as significant identifY significant normally will not constitute a separate separate violation. However, the circumstances surrounding circumstances surrounding the failure to correct may be considered considered relevant relevant to the determination determination of enforcement action for the initial inaccurateinaccurate or orincomplete incomplete statement. For example, example, an unintentionally inaccurate or incomplete unintentionally inaccurate incomplete submission may be treated as a more severe matter if the licensee later determines determines that the initial submittal was in error and does not correct it or if there were opportunities to identify were clear opportunities identifY the error. If If information information not corrected corrected was recognized by a licensee as significant, significant, a separate separate citation may be made for the failure to provide significant significant information. In any event, in serious cases where the licensee's licensee's actions in not correcting correcting oror providing information information raise questions about its commitment commitment to safety or its fundamental fundamental trustworthiness, trustworthiness, the Commission may exercise its authority to issue orders modifying, modifYing, suspending, or revoking the license. The Commission recognizes recognizes that enforcement enforcement determinations determinations must be made on a case-by-case case-by-case basis, taking into consideration consideration the issues described described in this section.

X. ENFORCEMENT ACTION ACTION AGAINSTAGAINST NON-LICENSEES NON-LICENSEES The Commission's Commission's enforcement enforcement policy is also applicable non-licensees, including applicable to non-licensees, including contractors contractors and subcontractors, holders of NRC approvals, e.g., certificates e.g., certificates of compliance, compliance, early site permits, standard standard design certificates, certificates, quality assurance program approvals, or applicants for quality assurance for any of them, and to employees of any of the foregoing, who knowingly knowingly provide components, equipment, equipment, or other goods or services services that relate to a licensee's activities subject to NRC regulation. The prohibitions prohibitions and sanctions sanctions for any of these persons who engage engage in deliberate deliberate misconduct misconduct or knowing knowing submission of incomplete incomplete or inaccurate inaccurate information are provided in the rule on deliberate misconduct, misconduct, e.g.,

e.g., 10 CFR 30.10 and 50.5. 50.5.

Contractors who supply products or services provided for use in nuclear activities are certain requirements subject to certain requirements designed to ensure ensure that the products products or services supplied that could affect safety are of high quality. Through procurement procurement contracts with licensees, licensees, suppliers may be required required to have quality assurance programs programs that meet applicable requirements, applicable requirements, e.g.,

Appendix B, and 10 CFR Part 71, 10 CFR Part 50, Appendix Contractors supplying certain 71, Subpart H. Contractors products products or services to licensees licensees are subject subject to the requirements of 10 CFR Part 21 regarding regarding reporting of defects in basic components.

. When inspections determine When inspections determine that that violations violations of of NRC NRC requirements requirements have occurred, or that contractors contractors have failed to fulfill contractual commitments (e.g.,

commitments (e.g., 10 CFR Part 50, Appendix B) that could adversely adversely affect the quality of a safety significant product or service, service, enforcement enforcement action action will be taken. Notices Notices of Violation and civil penalties will be used, as appropriate, for for contractors have programs licensee failures to ensure that their contractors programs that meet applicable requirements.

44

Notices of Violation will be issued for contractors contractors who violate 10 CFR Part 21. penalties

21. Civil penalties will be imposed imposed against individual directors directors or responsible officers of a contractor organization organization who knowingly and consciously consciously fail to provide provide the notice required by 10 CFR 21.21 (d)(1).

21.21(d)(1).

Notices of Violation or orders orders will be used against non-licensees non-licensees who are subject to the specific requirements requirements of Parts 71 and 72. Notices of Nonconformance Nonconformance will be used for contractors who fail to meet commitments related to NRC activities but are not in violation of specific specific requirements.

requirements.

XI. REFERRALS REFERRALS TO THE DEPARTMENT DEPARTMENT OF JUSTICE Alleged Alleged or suspected suspected criminal violations of the Atomic Energy Energy Act (and (and of other relevant Federal Federal laws) are referred referred to the Department Department of Justice (DOJ) for investigation. Referral to the DOJ does not preclude preclude the NRC from taking other enforcement enforcement action under this policy.

However, enforcement enforcement actions will be coordinated coordinated with the DOJ in accordance accordance with the Memorandum Understanding between Memorandum of Understanding between the NRC and the DOJ, (53 FR 50317; December December 14,14, 1988).

1988).

XlI. PUBLIC DISCLOSURE XII. DISCLOSURE OF ENFORCEMENT ENFORCEMENT ACTIONS ACTIONS Enforcement actions and licensees' Enforcement licensees' responses, in accordance accordance with 10 CFR 2.790, are publicly available available for inspection. In addition, press releases are generally issued for orders and civil penalties penalties and are issued at the same time the order or proposed proposed imposition of the civil penalty is issued. In addition, press releases are usually issued when a proposed civil penaltypenalty is withdrawn or substantially substantially mitigated by some amount. Press releases releases are not normally issued for Notices Notices of Violation that are not accompanied accompanied by orders or proposed proposed civil penalties.

XIII. REOPENING REOPENING CLOSED ENFORCEMENT ENFORCEMENT ACTIONS If significant new information information is received received or obtained by NRC which indicates indicates that an enforcement sanction enforcement sanction was incorrectly applied, consideration consideration may be given, dependent on the circumstances, to reopening a closed enforcement circumstances, enforcement action action to increase or decrease the severity severity of a sanction or to correct the record. Reopening Reopening decisions will be made on a case-by-case case-by-case basis, are expected to occur rarely, and require expected require the specific approval of the Deputy Executive ofthe Executive Director.

SUPPLEMENTS - VIOLATION SUPPLEMENTS VIOLATION EXAMPLESEXAMPLES This section provides examples of violations in each of four severity levels as guidance in determining appropriate severity level for violations in each of eight activity areas (reactor determining the appropriate (reactor operations, Part 50 facility construction, safeguards, operations, safeguards, health physics, transportation, fuel cycle and materials operations, miscellaneous miscellaneous matters, and emergency emergency preparedness).

preparedness).

SUPPLEMENT I--REACTOR SUPPLEMENT I--REACTOR OPERATIONS OPERATIONS 45

supplement provides This supplement provides examples examples of violations violations in each of the four severity levels as guidance in determining determining the appropriate severity appropriate severity level for violations violations in the area of reactor reactor operations.

A. Severity Level I - Violations Violations involving involving for example:

1.

1. A Safety Limit, as defined in 10 CFR 50.36 and the Technical Specifications Technical Specifications being exceeded; 2.

2. system'l22 designed A system designed to prevent or mitigate a serious safety event not being able to perform its intended safety function'function 133 when actually actually called upon to work; 3.

3. An accidental accidental criticality; or or

4. A licensed licensed operator operator at the controls of a nuclear nuclear reactor, or a senior operator operator directing licensed activities, involved in proceduralprocedural errors which which result in, or exacerbate exacerbate the consequences of, consequences of, an alert or higher level emergency and who, as a result of subsequent subsequent testing, receives a confirmed positive test result for drugs or alcohol.

B.

B. Severity Level H Il - Violations Violations involving for example:example:

1.

1. A system designed designed to prevent prevent or mitigate serious safety events not being able to to perform perform its intended safety function;

2. A licensed operator operator involved in the use, sale, or possession possession of illegal drugs or the consumption consumption of alcoholic beverages, within within the protected protected area; or 3.

3. A licensed licensed operator operator at the control of a nuclear nuclear reactor, or a senior operator operator directing licensed activities, involved in proceduralprocedural errors and who, as a result of subsequent subsequent testing, receives a confirmed positive test result for drugs or alcohol.

C. Severity Level III - Violations involving for example:

1.

1. A significant significant failure to comply with the Action Statement for a Technical Specification Specification Limiting Condition for Operation where where the appropriate actionaction was not taken within the required time, such as:

(a) In

[n a pressurized pressurized water reactor, in the applicable applicable modes, having one high-pressure safety injection pump inoperable for a period in excess of that allowed by the action statement; statement; or or 2

The term "system" as used in these supplements, 12The supplements, includes administrative administrative and managerial managerial control systems, as well as physical physical systems.

3

" "Intended safety l3"Intended safety function" means the function" means the total total safety function, and is not directed directed toward toward a loss of redundancy. A loss of one subsystem subsystem does not defeat the intended safety safety function as long as the other subsystem is operable.

46

(b) In a boiling water reactor, one primary containment isolation valve inoperable inoperable for a period in excess of that allowed by the action statement.

statement.

2. A system designed designed to prevent or mitigate a serious safety event not being able to perform its intended function under certain certain conditions (e.g.,

(e.g., safety system not operable operable unless offsite offsite power is available; materials or components available; materials components not environmentally environmentally qualified).

3.

3. Inattentiveness Inattentiveness to duty on the part of licensed personnel;

4. Changes in reactor parameters Changes parameters that cause unanticipated unanticipated reductions reductions in margins of of safety; 5.

5. A non-willful compromise of an application, application, test, or examination required required by 10 CFR Part 55 that:

(a) In the case of initial operator contributes to an individual operator licensing, contributes individual being granted an operator operator or a senior operator operator license, license, or (b) In the case of requalification, contributes contributes to an individual beingbeing permitted to perform the functions of an operator operator or a senior operator.

6. A licensee failure to conduct adequate oversight of contractors contractors resulting in the use of products or services that are of defective or indeterminate indeterminate quality and that have safety safety significance; significance;

7. A licensed operator's operator's confirmed positive positive test for drugs or alcohol that does not result in a Severity Severity Level I or II violation;

8. Equipment Equipment failures causedcaused by inadequate inadequate or improper improper maintenance maintenance that substantially complicates recovery substantially complicates recovery from a plant transient;

9. A failure to obtain obtain prior Commission Commission approval approval required by 10 CFR 50.59 for a

change, change, in which the consequence consequence of the change, is evaluated evaluated as having having low to moderate, or or greater greater safety safety significance (i.e., white, yellow, or red) by the SDP; significance (i.e.,

10. The failure to update update the FSAR FSAR as required by 10 CFR 50.71 50.7 1(e)

( e) where the unupdated unupdated FSARFSAR was used in performing performing a 10 CFR 50.59 evaluation for a change to the facility or procedures, implemented procedures, implemented without prior Commission Commission approval, that results in a condition condition evaluated evaluated as having low to moderate, significance (i.e.,

moderate, or greater safety significance (i.e., white, yellow, or red) by the SDP; or or 11.

11. The failure to make a report required by 10 CFR 50.72 or 50.73 associated with with any Severity Severity Level Level III violation.

47

D.

D. Severity Level IV - Violations Level IV- involving for example:

Violations involving example:

1. A less A less significant significant failure failure to to comply with the comply with the Action Statement for a Technical Action Statement Technical Specification Limiting Specification Limiting Condition Condition for Operation Operation where where the appropriate appropriate action action was was not taken taken within within the required time, such as:

the required (a) pressurized water In a pressurized water reactor, a 55 percent deficiency in percent deficiency in the required volume of the required volume condensate storage tank; or condensate storage or (b) In a boiling water subsystem of water reactor, one subsystem of the two independent MSIV leakage independent MSIV leakage subsystems inoperable; control subsystems inoperable;

2. A non-willful examination required compromise of an application, test, or examination non-willful compromise required by 10 CFR Part 55 that:

(a) In the case case of initial operator licensing, is discovered and reported to the NRC operator licensing, before an individual before granted an operator individual is granted operator or a senior senior operator operator license, oror (b) In the case discovered and reported requalification, is discovered case of requalification, before an reported to the NRC before individual is permitted perform the functions of an operator or a senior operator, or permitted to perform or (c)

(c) Constitutes more than minor concern.

3.

3. A failure to meet regulatory requirements that have more than minor safety regulatory requirements safety or or environmental significance; environmental significance;

4. A failure to make a required Licensee Event Report; required Licensee

5. Violations of 10 CFR 50.59 conditions evaluated as having very 50.59 that result in conditions very low low safety significance (i.e., green) by the SDP; or significance (i.e., or

6. A failure to update the FSAR as required by 10 CFR 50.71(e) in cases where the erroneous information is not used to make an unacceptable change to the facility or procedures.

erroneous E. Minor - Violations involving for example:

Minor A failure to meet 10 CFR 50.59 requirements requirements where there was not a reasonable likelihood reasonable likelihood that the change requiring 10 CFR 50.59 evaluation would ever require Commission review and approval prior to implementation. In the case ofa of a 10 10 CFR 50.71(e) violation, where a failure to update the FSAR would not have a material impact on safety or licensed activities.

48

SUPPLEMENT SUPPLEMENT lI--PART II--PART 50 FACILITY CONSTRUCTION CONSTRUCTION This supplement provides examples examples of violations in each each of the four severity severity levels as guidance in determining determining the appropriate appropriate severity level for violations in the area of Part 50 facility facility construction.

A. Severity Level LevellI - Violations Violations involving involving structures completed14 structures or systems that are completed 14 in such a manner that they would not have have satisfied their intended intended safety related purpose.

B. Severity Level II H - Violations involving for example:

1.

1. A breakdown breakdown in the Quality Assurance exemplified by Assurance (QA) program as exemplified deficiencies in construction QA related to more than one work activity (e.g.,

deficiencies (e.g., structural, piping, electrical, foundations). These deficiencies deficiencies normally involve the licensee's failure to conduct adequate adequate audits or to take prompt corrective corrective action on the basis of such audits and normally involve multiple examples examples of deficient deficient construction construction or construction construction of unknown unknown quality due to inadequate program program implementation; or or

2. A structure or system that is completed completed in such a manner that it could could have an adverse effect on the safety of operations.

operations.

C. Severity Level IIIIII - Violations Violations involving involving for example:

1. A deficiency deficiency in a licensee QA program program for construction construction related to a single work (e.g., structural, piping, electrical, or foundations). This significant deficiency normally activity (e.g.,

involves the licensee's failure to conduct adequate audits or to take prompt corrective corrective action on on the basis of such audits, and normally normally involves involves multiple examples examples of deficient deficient construction or construction or construction of unknown unknown quality due to inadequate program implementation; implementation;

2. A failure to confirm the design safety safety requirements requirements of a structure or system as a result of inadequate preoperational preoperational test program implementation; implementation; or or 3.

3. A A failure to make a required 10 CFR 50.55(e) report.

D. Severity Level IV IV - Violations Violations involving failure to meet regulatory requirements requirements including one or more Quality Assurance Criterion not amounting to Severity Severity Level I,I, II, or III violations that have more than minor safety or environmental environmental significance.

"4The term "completed" 14The "completed" as used in this supplement supplement means means completion of construction construction including review review and and acceptance construction QA organization.

acceptance by the construction 49

SUPPLEMENT SUPPLEMENT Ill--SAFEGUARDS IH--SAFEGUARDS supplement provides examples of violations This supplement violations in each of the four severity levels as guidance in determining determining the appropriate appropriate severity level for violations in the area of safeguards.

A. Severity Level I - Violations involvinginvolving for example:

1. An act of radiological radiological sabotage in which the security security system did not function as required and, as a result of the failure, there was a significant event, such as:

(a)

(a) A Safety Limit, as defined in 10 CFR 50.36 and the Technical Technical Specifications, Specifications, was exceeded; (b) A system designed to prevent or mitigate a serious serious safety event was not able to to perform its intended safety function when actually called upon to work; or or (c)

(c) An accidental accidental criticality occurred;

2. The theft, loss, or diversion quantity15 of special nuclear material diversion of a formula quantityl5 material (SNM); or or 3.

3. Actual unauthorized production Actual unauthorized production of a formula quantity of SNM.

B. Severity Level H II - Violations involving for example: example:

1.

1. The entry of an unauthorized individual l66 who represents unauthorized individual' represents a threat area17 threat into a vital areal?

from outside the protected protected area;

2. The theft, loss or diversion ofSNM significance"l88 in which of SNM of moderate strategic significance the security system did not function as required; required; or or 3.

3. Actual Actual unauthorized unauthorized production of SNM.

ofSNM.

"See 10 CFR 73.2 15See 73.2 for the definition of"formula of "formula quantity."

16The term 16The "unauthorized individual" term "unauthorized individual" asas used used in this supplement in this supplement means means someone someone who was not authorized authorized for entrance into the area in question, or not authorized to enter authorized enter in the manner entered.

17 The phrase 17The "vital area" phrase "vital as used area" as used in this supplement in this supplement includes includes vital vital areas areas and material access and material access areas.

8 See lO 18See 10 CFR 73.2 for the definition ofof"special "special nuclear material of moderate moderate strategic significance."

significance."

50 50

c.

C. Severity Level III - Violations Violations involving involving for example:

1. A failure or inability to control control access access through established systems or procedures, procedures, such that an unauthorized unauthorized individual (i.e.,

(i.e., not authorized unescorted unescorted access to protected protected area) could easily access'9l9 into a vital area from outside the protected easily gain undetected access protected area;

2. A failure to conduct conduct any search at the access control point or conducting conducting an inadequate inadequate search that resulted in the introduction introduction to the protected protected area of firearms, explosives, explosives, oror incendiary incendiary devices and reasonable facsimiles thereof that could significantly assist radiological radiological sabotage or theft of strategic sabotage strategic SNM; 3.

3. A failure, degradation, degradation, or other deficiency of the protected protected area intrusion detection detection or alarm assessment assessment systems such that an unauthorized unauthorized individual who represents represents a threat threat could predictably circumvent predictably circumvent the system or defeat a specific zone with a high degree of confidence without insider insider knowledge, or other significant significant degradation of overall system capability;

4. A significant failure ofthe of the safeguards safeguards systems designed or used to prevent or or detect the theft, loss, or diversion of strategic strategic SNM; 5.

5. A failure to protect or control classified safeguards information considered classified or safeguards considered to be significant while the information is outside the protected protected area and accessible accessible to those not not authorized authorized access to the protected protected area;

6. A significant significant failure to respond to an event either in sufficient time to provide provide protection protection to vital equipment equipment or strategic SNM, or with an adequate adequate response force; or or

7. A failure to perform appropriate evaluation perform an appropriate evaluation or background investigation so that background investigation information information relevant to the access determination determination was not obtained obtained or considered considered and as a result a person, who would likely not have been granted granted access by the licensee, licensee, if the required required investigation or evaluation had been performed, investigation performed, was granted access.

D. Severity Level Level IV IV - Violations involving involving for example:

1.

1. A failure or inability to control access such that an unauthorized individual (i.e.,

unauthorized individual (i.e.,

authorized protected area but not to vital area) could easily authorized to protected easily gain undetected access access into a vital area from inside the protected protected area or into a controlled access controlled access area;

2. A failure to respond to a suspected suspected event in either either a timely manner or with an an adequate adequate response response force; 9In "19In determining whether access can be easily gained, factors such as predictability, identifiability, and ease of passage should be considered.

51

3.

3. A failure to implement 10 CFR Parts 25 and 95 with respect to the information information addressed under Section 142 142 of the Act, and the NRC approved security security plan relevant to those parts;

4. A failure to conduct conduct a proper search at the access control point;

5. A failure to properly properly secure or protect classified classified or safeguards safeguards information inside the protected protected area that could could assist an individual in an act of radiological sabotage or theft of of strategic SNM where the informationinformation was not removed removed from the protected protected area;

6. A failure to control access such that an opportunity exists that could allow allow unauthorized unauthorized and undetected undetected access into the protected protected area but that was neither easily or likely to to be exploitable;

,7.

I 7. A failure to conduct an adequate adequate search search at the exit from a material material access area;

8. A theft or loss of SNM oflow of low strategic significance significance that was not detected within the time period specified in the security security plan, other relevant document, or regulation; regulation; or

9. Other Other violations that have more than minor safeguards safeguards significance.

SUPPLEMENT IV--HEALTH IV--HEALTH PHYSICS PHYSICS (10 (10 CFR PART 20)

This supplement supplement provides examples of violationsviolations in each of the four severity levels as guidance in determining guidance determining the appropriate appropriate severity level for violations in the area of health physics, 20 10 CFR Part 20. 20.20 A. Severity Level I - Violations Violations involving for example:

1.

1. A radiation exposure during any year of a worker worker in excess of 25 rems reins total effective dose equivalent, equivalent, 75 remsrerns to the lens ofthe of the eye, or 250 rads to the skin of the whole ofthe body, or to the feet, ankles, hands or forearms, or to any other organ or tissue;

2. A radiation exposure over the gestation period of the embryo/fetus declared embryo/fetus of a declared pregnant pregnant woman in excess of2.5 of 2.5 rems total effective dose equivalent;

3. A radiation exposure during any year ofa of a minor in excess of2.5 of 2.5 rems total effective dose equivalent, equivalent, 7.5 rems to the lens of the eye, or 25 rems rerns to the skin of the whole body, or to the feet, ankles, hands or forearms, or to any other organ organ or tissue; 2°Personnel overexposures overexposures and associated 2oPersonnel associated violations incurred during a life-saving life-saving or other emergency emergency response effort will be treated on a case-by-case case-by-case basis.

52 52

4. An annual exposure of a member of the public in excess of 1.0 rem total effective dose equivalent; equivalent;

5. A release of radioactive radioactive material material to an unrestricted unrestricted area at concentrations concentrations in excess of 50 times the limits for members members of the public as described 20.1302(b)(2)(i);

described in 10 CFR 20.1302(b )(2)(i);

or or

6. Disposal Disposal of licensed material in quantities or concentrations concentrations in excess of 10 times the limits of 10 CFR 20.2003.

20.2003.

B.

B. Severity Level H II - Violations Violations involving for example:

example:

1.

1. A radiation exposure exposure during any year of a worker in excess excess of 10 rems total effective effective dose equivalent, 30 rems rerns to the lens of the eye, or 100 rems rerns to the skin of the whole body, or to the feet, ankles, hands or forearms, or to any other organ or tissue;

2. A radiation exposure exposure over the gestation period of the embryo/fetus of a declared declared pregnant woman in excess of 1.0 rem total effective dose equivalent; equivalent; 3.

3. A A radiation exposure exposure during any year of a minor in excess of 1 rem total effective effective dose equivalent; rems to the lens of the eye, or 10 rems to the skin of the whole body, or to equivalent; 3.0 rerns to the feet, ankles, hands or forearms, or to any other organ or tissue;

4. An annual exposure of ofaa member of the public public in excess of 0.5 rem total effective effective dose equivalent; equivalent; 5.

5. A release of radioactive radioactive material to an unrestricted unrestricted area at concentrations concentrations in excess excess of 10 times the limits for members members of the public as described 20.1302(b)(2)(i) described in 10 CFR 20.1302(b )(2)(i)

(except (except when operation operation up to 0.5 rem a year has been approved by the Commission under

§20.1301(c));

§20.1301(c));

6. Disposal of licensed material in quantities or concentrations concentrations in excess of five times the limits of 10 20.2003; or 10 CFR 20.2003; or

7. A failure to make an immediate immediate notification notification as required by 10 CFR 20.2202 (a)(1)(a)(1) or (a)(2).

(a)(2).

c.

C. Severity Level III - Violations involving for example:

example:

1.

1. A radiation exposure exposure during any year of a worker in excess excess of 5 rems total effective effective dose equivalent, 15 rems rerns to the lens of the eye, or 50 rems reins to the skin of the whole body or to the feet, ankles, hands or forearms, or to any other organ or tissue; 53 53

2. A radiation exposure over the gestation period of the embryo/fetus embryo/fetus of a declared declared pregnant woman in excess of 0.5 rem total effective effective dose equivalent equivalent (except (except when doses are in accordance accordance with the provisions of §20.1208(

§20.1208(d));

d));

3.

3. A radiation radiation exposure during any year ofa of a minor minor in excess excess of 0.5 rem total effective dose equivalent; equivalent; 1.5 rerns rems to the lens of the eye, or 5 rems to the skin of the whole body, or to the feet, ankles, ankles, hands or forearms, or to any other organ or tissue;

4. An annual exposure exposure of a member of the public in excess excess of 0.1 rem total effective effective dose equivalent equivalent (except (except when operation operation up to 0.5 rem a year has been approved be.en approved by the Commission under §20.1301(c));

Commission §20.1301(c));

5. A release of radioactive radioactive material to an unrestricted unrestricted area at concentrations concentrations in excess of two times the effluent effluent concentration concentration limits referenced referenced in 1100 CFR 20.1302(b 20.1302(b)(2)(i)

)(2)(i)

(except when operation up to 0.5 rem a year has been approved approved by the Com1nission Commission under under 2 0.1301(c));

Section 20.1301(c));

Section

6. A failure to make a 24-hour 24-hour notification required by 10 CFR 20.2202(b) or an an immediate immediate notification required by 10 CFR 20.2201(a)(1)(i);

20.2201(a)(1)(i);

7. A substantial substantial potential potential for exposures exposures or releases releases in excess of the applicable applicable limits limits 20.1001-20.2401 whether or not an exposure in 10 CFR 20.1001-20.2401 exposure or release occurs;

8. Disposal of licensed material not covered covered in Severity Severity Levels I1 or II;

9. A release for unrestricted unrestricted use of contaminated contaminated or radioactive radioactive material oror equipment that poses a realistic potential for exposure of the public to levels or doses exceeding exceeding the annual dose limits for members of the public;

10. of licensee activities by a technically Conduct oflicensee technically unqualified person; or 11.

11. A violation involving failure to secure, or maintain surveillance surveillance over, licensed licensed material that:

(a) involves licensed material in any aggregate quantity greater than 1000 times the quantity specified in Appendix Appendix C to Part 20; or or (b) involves licensed licensed material in any aggregate aggregate quantity greater greater than 10 times the quantity quantity specified in Appendix C to Part 20, where such failure is accompanied specified accompanied by the absence of of a functional program to detectdetect and deter security violations that includes training, staff staff (including auditing), and corrective awareness, detection (including corrective action (including (including disciplinary action); oror 54 54

(c) results in a substantial potential (c) potential for exposures or releases in excess of the applicable limits in Part 20.

D. Severity Level Level IV- Violations involving for example:

IV - Violations example:

1.

1. Exposures in excess of the limits of 10 CFR 20.1201,20.1201, 20.1207, 20.1207, or 20.1208 20.1208 not not constituting constituting Severity Level I, II, or III violations; Severity Level violations;

2. A release of radioactive radioactive material to an unrestricted unrestricted area at concentrations in in excess excess of the limits for members members of the public as referenced referenced in 10 CFR 20. 20.1302(b)(2)(i) 1302(b)(2)(i) (except (except when operation operation up to 0.5 rem a year has been approved by the Commission under §20.130l(c)); §20.1301(c));

3.

3. A radiation radiation dose rate in an unrestricted unrestricted or controlled controlled area in excess of 0.002 rem in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (2 millirem/hour) milhirem/hour) or 50 millirems in a year;

4. Failure to maintain and implement radiation radiation programs programs to keep radiation exposures exposures as low as is reasonably reasonably achievable; achievable;

5. Doses to a member of the public in excess of any EPA generally applicable environmental radiation standards, such as 40 CFR Part 190; environmental 190;

6. A failure to make the 30-day notification notification required required by 10 CFR 20.2201(a)(1)(ii) 20.220l(a)(1)(ii) or or 20.2203(a);

, 7. A failure to make a timely written report as required required by 10 CFR 20.2201(b),

20.220l(b),

20.2204, or 20.2206; 8.

8. A.

A failure to report an exceedance exceedance of the dose constraint established established in 10 CFR 20.1101 20.110 1(d) or a failure to take corrective corrective action action for an exceedance, exceedance, as required required by by 10 CFR 20.1101(d);

20.1101(d);

9. Any other matter that has more than a minor minor safety, health, or environmental environmental significance; or significance; or

10. A violation violation involving an isolated failure to secure, or maintain surveillance over, licensed material that is not otherwise characterized in Example IV.C.ll otherwise characterized IV.C. 11 and that involves licensed licensed material in any aggregate aggregate quantity greater than 10 times the quantity specified in Appendix C to Part 20, provided that: (i) the material is labeled as radioactive radioactive or located in an containing radioactive materials; area posted as containing (ii) such failure occurs despite a functional materials; and (ii) program to detect and deter security violations that includes training, staff awareness, detectiondetection corrective action (including (including auditing), and corrective (including disciplinary action).

55

E. Minor - Violations involving for example:

Minor example:

A A violation involving an isolated failure to secure, or maintain maintain surveillance over, licensed licensed material in an aggregate quantity quantity that does not exceed exceed 10 times the quantity specified in Appendix C to Part 20.

SUPPLEMENT V--TRANSPORTATION SUPPLEMENT V--TRANSPORTATION supplement provides examples This supplement examples of violations in each each of the four severity levels as ofthe guidance in determining determining the appropriate severity level for violations in the area of NRC transportation requirements. 221 transportation "

A. Severity Levell Level I - Violations involving for example: example:

I.

]. Failure to meet transportation transportation requirements requirements that resulted control of resulted in loss of control of radioactive material material with a breach in package package integrity such that the material material caused a radiation radiation exposure to a member of the public and there was clear potential potential for the public to receive more than. .11 rem to the whole than whole body;

2. contamination in excess of 50 times the NRC limit; or Surface contamination or 3.

3. External radiation levels in excess of 10 times the NRC NRC limit.

B. Severity Level H II - Violations involving involving for example:

I. Failure Failure to meet transportation requirements that resulted in loss of control of of radioactive radioactive material with a breach in package package integrity such that there was a clear potential for for the member member ofthe of the public to receive more than than. .11 rem to the whole body;

2. contamination in excess of 10, but not more than 50 times the NRC limit; Surface contamination

3. External External radiation levels in excess offive, of five, but not more more than 10 times the NRC limit; or or

4. A failure to make required initial notifications associated with Severity notifications associated Level I or Severity Levell or 11 II violations.

C. Severity Level III - Violations involving for example: example:

2

'Some transportation 21Some requirements are transportation requirements applied to are applied to more than one licensee involved in the same activity such as a more than shipper and a carrier. When a violation violation of such a requirement occurs, enforcement enforcement action will be directed directed against against the responsible licensee which, under the circumstances circumstances of the case, may be one or more more of the licensees involved.

56 56

1.

1. Surface contamination in excess of five but not more than 10 times the NRC limit; Surface

2. External radiation radiation in excess of one but not more than five times the NRC limit;

3. noncompliance with labeling, placarding, shipping paper, packaging, loading, Any noncompliance or other requirements requirements that could reasonably reasonably result in the following:

(a) A significant failure to identify the type, quantity, or form of material; (b) of the carrier or recipient A failure ofthe recipient to exercise adequate adequate controls; oror (c)

(c) A substantial potential for either personnel personnel exposure or contamination contamination above regulatory limits or improper transfer of material; or or

4. A failure to make required required initial notification associated associated with Severity Level III III violations.

vio lations.

D. Severity Level IV - Violations involving involving for example:

1.

1. A breach of package package integrity without without external radiation exceeding the radiation levels exceeding NRC limit or without contamination levels exceeding exceeding five times the NRC limits;

2. contamination in excess of but not more than five times the NRC limit; Surface contamination 3.

3. A failure to register as an authorized authorized user of an NRC-Certified NRC-Certified Transport package; package;

4. A noncompliance noncompliance with shipping papers, marking, labeling, placarding, placarding, packaging packaging or loading not amounting amounting to a Severity Level I, 1I,II, or III violation; 5.

5. A failure to demonstrate demonstrate that packages for special special form radioactive material meets applicable regulatory requirements; applicable regulatory requirements;

6. A failure to demonstrate demonstrate that packages packages meet DOT Specifications Specifications for 7A 7A Type A packages; packages; or 7.

7. Other violations that have more than minor safety or environmental environmental significance.

SUPPLEMENT SUPPLEMENT Vt--FUEL VI--FUEL CYCLE AND AND MATERIALS MATERIALS OPERATIONS OPERATIONS examples of violations in each of the four severity levels as This supplement provides examples guidance guidance in determining the appropriate severity level for violations in the area of fuel cycle and materials operations.

57 57

A. Severity Level I - Violations involving for example: example:

1.

1. Radiation Radiation levels, contamination contamination levels, or releases releases that exceed 10 times the limits limits specified in the specified license; the license;

2. A system designed designed to prevent or mitigate a serious safety event not being operable when actually required required to perform perform its design function; 3.

3. A nuclear nuclear criticality criticality accident;

4. Failure to use a properly properly prepared written directive directive as required required by 10 CFR 35.40; or failure to develop, implement, implement, or maintain procedures for administrations maintain procedures administrations requiring requiring a written directive as required by 10 CFR 35.41; that results in a death or serious injury (e.g., (e.g., substantial organ impairment);

impairment);

5. A safety limit, as defined in 10 CFR 76.4, the Technical Safety Requirements, Requirements, or or the application application being exceeded; exceeded; or or

6. Significant Significant injury or loss oflife of life due to a loss of control over licensed or certified certified activities, including activities, including chemical processes processes that are integral to the licensed or certified certified activity, whether radioactive radioactive material is released or not.

B.

B. Severity Level II HI-- Violations involving for example:

example:

1. Radiation levels, contamination contamination levels, or releases that exceed five times the limits specified specified in the license;

2. A system designed to prevent or mitigate a serious safety event being inoperable; 3.

3. programmatic failure to implement written A substantial programmatic written directives directives or procedures procedures for administrations administrations requiring requiring a written directive, such as a failure ofthe of the licensee's procedures procedures to address one or more of the elementselements in 10 CFR 35.40 or.35.41, or a failure to train personnel 35.40 or-35.41, personnel in those procedures, that results in a medical medical event;

4. A failure to establish, implement, or maintain all criticality controls (or control control systems) systems) for a single nuclear criticality criticality scenario when a critical critical mass of fissile material was material present or reasonably available, such that a nuclear reasonably available, nuclear criticality criticality accident was possible; or or 5.

5. The potential for a significant significant injury or loss of life due to a loss of control over licensed licensed or certified activities, including chemical processes processes that are integral to the licensed licensed or or certified certified activity, whether whether radioactive material is released or not (e.g., (e.g., movement movement of liquid UF UF 66 cylinder cylinder by unapproved unapproved methods).

58 58

C. Severity Level III - Violations involving for example: example:

1. Possession Possession or use of unauthorized unauthorized equipment or materials materials in the conduct of of licensee activities which degrades safety; licensee

2. Use of radioactive material on humans where such use is not authorized; 3.

3. Conduct of licensed activities by a technically unqualified Conduct unqualified or uncertified uncertified person;

4. A substantial substantial potential potential for exposures, radiation contamination levels, radiation levels, contamination levels, or or

releases, releases, including releases releases of toxic material caused caused by a failure to comply with NRC regulations, licensed or certified activities in excess of regulatory limits; from licensed

5. A substantial substantial programmatic programmatic failure to implement written directives or procedures procedures for administrations administrations requiring a written directive, such as a failure of the licensee's licensee's procedures to procedures address address one or more of the elements in 10 CFR 35.40 or 35.41, 35.41, or a failure to train personnel in those procedures, that does not result in a medical medical event. Failure to report a medical event. A programmatic programmatic weakness implementation of written directives or procedures weakness in the implementation procedures for administrations requiring administrations requiring a written directive, directive, whether or not a medical event occurs;

6. A failure, during radiographic radiographic operations, operations, to have present present at least two qualified qualified individuals or to use radiographic radiographic equipment, equipment, radiation survey survey instruments, and/or personnel monitoring monitoring devices as required required by 10 CFR Part 34;

7. A failure to submit an NRC Form 241 as required by 10 CFR 150.20;

8. A failure to receive receive required required NRC approval priorprior to the implementation implementation of a change in licensed activities that has radiological programmatic significance, radiological or programmatic significance, such as, a change in ownership; ownership; lack of an RSO or replacement of an RSO with an unqualified individual; individual; a change in the location location where where licensed activities are being conducted, or where licensed licensed material material is being being stored where the new facilities do not meet the safety guidelines; or a change in the quantity or or type of radioactive processed or used that has radiological radioactive material being processed radiological significance;

9. A significant significant failure to meet decommissioning requirements requirements including a failure to to notify the NRC as required by regulation notifY regulation or license condition, substantial substantial failure to meet decommissioning standards, decommissioning standards, failure to conduct and/or complete decommissioning activities complete decommissioning activities in accordance with regulation or license condition, or failure to meet required accordance required schedules without without adequate justification;

10. A significant failure to comply with the action statement for a Technical Technical Safety Requirement Limiting Requirement Limiting Condition for Operation Operation where the appropriate appropriate action was not taken within the required required time, such as:

59 59

(a) In an autoclave, where a containment containment isolation valve is inoperable inoperable for a period in in excess of that allowed by the action statement; or action statement; or (b) Cranes or other lifting devices engaged in the movement of cylinders having inoperable inoperable safety components, such as redundantredundant braking systems, or other safety devices for a period in excess of that allowed by the action statement; 11.

11. A system designed to prevent or mitigate a serious safety event:

(a)

(a) Not being able to perform perform its intended intended function under certain conditions (e.g., (e.g., safety safety system not operable unless utilities available, available, materials or components not according to specifications); or specifications);

(b) Being degraded to the extent that a detailed evaluation would be required to detailed evaluation determine determine its operability; 12.

12. Changes Changes in parameters parameters that cause unanticipated unanticipated reductions in margins of safety; 13.

13. A significant failure to meet the requirements of 10 CFR 76.68, requirements oflO 76.68, including including a failure such that a required required certificate certificate amendment amendment was not sought;

14. of the certificate A failure ofthe certificate holder to conduct conduct adequate adequate oversight of contractors resulting in the use of products or services that are of defective or indeterminate indeterminate quality and that have safety significance; significance; 15.

15. Equipment failures caused by inadequate or improper maintenancemaintenance that substantially complicates recovery from a plant transient; substantially 16.

16. A failure to establish, maintain, or implement all but one criticality A (or criticality control (or control systems) for a single nuclear criticality scenario when a critical mass offissile nuclear criticality of fissile material was present present or reasonably available, available, such that a nuclear criticality accident accident was possible; or or 17.

17. A failure, during radiographic radiographic operations, to stop work after a pocket dosimeterdosimeter is is found to have gone off-scale, off-scale, or after an electronic electronic dosimeter reads greater greater than 200 mrem, and before a determination determination is made of the individual's actual radiation exposure.

D. Severity Level IV IV-- Violations involving for example:

1.

l. A failure to maintain maintain patients hospitalized hospitalized who have cobalt-60, cesium-137, cesium-137, or or iridium- 192 iridium-192 implants or to conduct required leakage or contamination contamination tests, or to use properly calibrated equipment; calibrated 60

2. Other violations violations that have more than minor safety or environmental environmental significance; significance;

3. Failure to use a properly properly prepared written directivedirective as required required by 10 CFR 35.40; or or failure to develop, implement, or maintain procedures procedures for administrations requiring a written directive directive as required by 10 CFR 35.41, 35.41, whether or not a medicalmedical event occurs, provided provided that the failures: (1) are isolated; (2) do not demonstrate programmatic weaknesses in demonstrate programmatic weaknesses implementation; implementation; and (3)

(3) have limited consequences consequences if a medical medical event is involved;

4. A failure to keep the records required required by 10 CFR 35.32 or 35.33; 35.33;

5. A less significant significant failure to comply with the Action Statement Statement for a Technical Safety Requirement Requirement Limiting Condition Condition for Operation when the appropriate appropriate action was not taken within the required required time;

6. A failure to meet the requirements of 10 CFR 76.68 that does not result in a Severity Severity Level 1, 11, Levell, II, or III violation;

7. A failure to make a required written event report, as required required by 10 CFR 76.120(d)(2);

76.120(d)(2);

or or

8. A failure to establish, implement, or maintain a criticality control control (or control control system) for a single nuclear criticality scenario scenario when the amount of fissile material available available was not, but could have been sufficient to result in a nuclear criticality.

SUPPLEMENT SUPPLEMENT VIl--MISCELLANEOUS VII--MISCELLANEOUS MATTERS This supplement provides examples examples of violations in each ofthe of the four severity levels as guidance in determining appropriate severity level for violations involving miscellaneous determining the appropriate miscellaneous matters.

A. Severity Level I - Violations involvinginvolving for example:

example:

1. Inaccurate information22 Inaccurate or incomplete information 22 that is provided provided to the NRC (a) deliberately deliberately with the knowledge of a licensee official that the information is incomplete incomplete or inaccurate, inaccurate, or (b) if the information, had it been complete and accurate at the time provided, likely would have resulted in regulatory regulatory action such as an immediateimmediate order required by the public health and safety; 22In 22In applying applying the the examples examples in this supplement in this supplement regarding regarding inaccurate inaccurate or incomplete information or incomplete information and and records, records, reference reference should also be made to the guidance guidance in Section IX, "Inaccurate "Inaccurate and Incomplete Incomplete Information,"

Information," and to the the definition of"licensee of "licensee official" contained contained in Section IV.C.

61

2. Incomplete or inaccurate Incomplete inaccurate information information that the NRC requires be kept by a licensee that is ((a) a) incomplete or inaccurate inaccurate because of falsification by or with the knowledge knowledge of a licensee official, or (b) if the information, had it been complete and accurate when reviewed by by the NRC, likely would have resulted in regulatory regulatory action such as an immediate order required by by public health and safety safety considerations; considerations;

3. Information that the licensee has identified as having significant implications Information implications for for public health and safety safety or the common common defense and security ("significant information identified security ("significant identified by a licensee")

licensee") and is deliberately deliberately withheld firom the Commission; withheld from

4. Action by senior corporate corporate management management in violation of 10 CFR 50.7 or similar regulations regulations against an employee; employee;

5. A knowing and intentional intentional failure to provide the notice required by 10 CFR Part 21; oror 23

6. implement the substantially implement A failure to substantially the required fitness-for-duty program.

required fitness-for-duty program. 23 B. Severity Level II H - Violations involving for example: example:

1. Inaccurate Inaccurate or incomplete information that is provided provided to the NRC (a) (a) by a licensee licensee official because of careless careless disregard for the completeness completeness or accuracy accuracy of the information, or (b) if the information, had it been completecomplete and accurate accurate at the time provided, likely would have resulted in regulatory regulatory action such as a show cause order or a different regulatory regulatory position;

2. Incomplete inaccurate information that the NRC requires be kept by a licensee Incomplete or inaccurate which is (a) incomplete incomplete or inaccurate because because of careless careless disregard for the accuracy of the information information on the part of a licensee official, or (b) if the information, had it been complete and accurate accurate when reviewed by the NRC, likely would have resulted in regulatory regulatory action such as a show cause order or a different regulatory position;

3. "Significant "Significant information information identified by a licensee" and not provided to the Commission because because of careless disregard disregard on the part part of a licensee licensee official;

4. An action by plant management management or mid-level management in violation of 10 CFR 50.7 or similar regulations regulations against an employee; 5.

5. A failure to provide the notice required required by 10 CFR Part 21; 21; 23The example for 23The example for violations violations for for fitness-for-duty fitness-for-duty relate relate to violations of to violations of 10 I 0 CFR Part 26.

CFR Part 62

6. A failure to remove an individual from unescorted unescorted access who has been involvedinvolved protected area or take action for on duty in the sale, use, or possession of illegal drugs within the protected misuse of alcohol, prescription drugs, or over-the-counter over-the-counter drugs;

7. .- A failure to take reasonable action when observedobserved behavior behavior within within the protected protected area or credible information concerning credible information concerning activities within the protected protected area indicates indicates possible unfitness unfitness for duty based on drug or alcohol use;

8. A deliberate deliberate failure of the licensee's Employee Assistance Assistance Program (EAP) to notify notifY licensee's licensee's management management when EAP's staff is aware that an individual'sindividual's condition condition may adversely affect affect safety safety related related activities; activities; or or

9. The failure oflicensee of licensee management management to take effective effective action action in correcting a hostile work environment.

C. Severity Level III - Violations Violations involving for example:

example:

1.

1. Incomplete Incomplete or inaccurate inaccurate information that is provided to the NRC (a) (a) because ofof inadequate actions on the part of licensee oflicensee officials but not amounting amounting to a Severity Level I or lor II violation, or (b) if the information, had it been complete accurate at the time provided, likely complete and accurate would have resulted in a reconsideration reconsideration of a regulatory position or substantial further inquiry such as an additional inspection or a formal request for information;

2. Incomplete or inaccurate Incomplete inaccurate information information that the NRC requires be kept by a licensee licensee that is (a) incomplete incomplete or inaccurate because because of inadequate inadequate actions on the part of oflicensee licensee officials but not amounting to a Severity Severity Level Level I or II violation, or (b) if the information, had it been been complete and accurate when reviewedreviewed by the NRC, likely would have have resulted in a reconsideration of a regulatory position or substantial further inquiry such as an additional reconsideration additional inspection or a formal request for information; 3.

3. Inaccurate Inaccurate or incomplete performance performance indicator (PI) data submitted to the NRC by a Part 50 licensee licensee that would have caused caused a PI to change change from green to either either yellow yellow or red; white to either yellow or red; or yellow to red.

4. A failure to provide provide "significant information identified identified by a licensee" licensee" to the Commission and not amounting amounting to a Severity Level I or II 11 violation;

5. An action by first-line supervision supervision or other low-level low-level management in violation of of 10 CFR 50.7 or similar regulations regulations against an employee;

6. An inadequate review or failure to review such that, if an appropriate appropriate review had had been made as required, a 10 CFR Part 21 report would have been made; 63

7. A failure to complete complete a suitable inquiry on the basis of 10 CFR Part 26, keep records records concerning concerning the denial denial of access, or respond to inquiries concerning denials of access so inquiries concerning so that, as a result of the failure, a person person previously previously denied access access for fitness-for-duty fitness-for-duty reasons was improperly granted access;

8. A failure to take the required action for a person confirmed to have been tested tested positive for illegal drug use or take action for onsite onsite alcohol use; not amounting to a Severity Level Level II violation;

9. A failure to assure, as required, that contractors have an effectiveeffective fitness-for-duty fitness-for-duty program; or program; or

10. Threats Threats of discrimination discrimination or restrictive agreements agreements which are violations under NRC regulations regulations such as 10 CFR 50.7(f.50.7(f).

D. Severity Level IV IV-- Violations Violations involving involving for example:

1.

1. Incomplete or inaccurate Incomplete inaccurate information information that is provided to the NRC but not not amounting to a Severity Level Level I, II, or III or.III violation;

2. Information Information that the NRC requires be kept by a licensee licensee and that is incomplete or inaccurate and of more than minor significance significance but not amounting to a Severity Level I, II, or III violation;

3. Inaccurate or incomplete Inaccurate incomplete performance indicator (PI) data submitted to the NRC by performance indicator by a Part 50 licensee that would have caused a PI to change from green to white.

4. An inadequate review or failure to review under 10 CFR Part 21 or other other procedural violations associated with 10 CFR Part 21 with more than minor safety significance;

5. Violations of the requirements requirements of Part 26 of more than minor significance;

6. A failure to report acts of licensed operators or supervisors pursuant pursuant to 10 CFR 26.73; or

7. Discrimination cases which, in themselves, do not warrant a Severity Level III categorization.

E. Minor Minor - Violations involving involving for example:

Inaccurate or incomplete incomplete performance performance indicator (PI) data submitted to the NRC by a Part 50 licensee that would not have caused caused a PI to change change color.

64 64

SUPPLEMENT SUPPLEMENT VHI--EMERGENCY PREPAREDNESS VIII--EMERGENCY PREPAREDNESS supplement provides This supplement provides examples of violations in each ofthe of the four severity levels as guidance in determining the appropriate appropriate severity severity level for violations violations in the area of emergency emergency preparedness. It should be noted preparedness. noted that citations are not normally made for violations violations involving emergency preparedness emergency preparedness occurring occurring during emergency emergency exercises. However, where exercises (i) training, procedural, or repetitive failures for which corrective reveal (i) corrective actions have not been taken, (ii)

(ii) an overall concern regarding regarding the licensee's ability to implement its plan in a manner that adequately protects public health and safety, or (iii) poor self critiques of the licensee's licensee's exercises, enforcement action may be appropriate.

exercises, enforcement appropriate.

A. Severity Level I - Violations involving for example:

In a general emergency, emergency, licensee licensee failure to promptly (1) correctly classifyclassify the event, (2)

(2) make required notifications notifications to responsible Federal, State, and local agencies, or (3)

(3) respond to to the event (e.g.,

(e.g., assess actual or potential offsite consequences, consequences, activate activate emergency emergency response facilities, and augment shift staff) staff)....

B. Severity Level H II - Violations Violations involving for example:

example:

1.

1. In a site emergency, licensee failure to promptly emergency, licensee promptly (1) correctly correctly classify the event, (2)

(2) make required notifications notifications to responsible Federal, State, and local agencies, or (3) (3) respond to to the event (e.g.,

(e.g., assess actual or potential off offsite consequences, activate site consequences, activate emergency emergency response facilities, facilities, and augment shift staff); or

2. A licensee failure to meet or implementimplement more than one emergency planning planning standard standard involving assessment assessment or notification.

C. Severity Level Level III - Violations involving for example: example:

1.

1. In an alert, licensee failure to promptly (1) correctly correctly classify the event, (2) (2) make required notifications required notifications to responsible Federal, State, Federal, State, and local agencies, agencies, or (3)

(3) respond to the event (e.g.,

(e.g., assess actual actual or potential offsite consequences, activate emergency response offsite consequences, response facilities, and augment augment shift staff); or or

2. A licensee licensee failure to meet or implement one emergency emergency planning planning standard involving assessment or notification.

D.

D. Severity Level Level IVIV - Violations involving for example:

A licensee licensee failure to meet or implement any emergency emergency planning planning standard standard or requirement requirement not directly related to assessment and notification.

65

INTERIM ENFORCEMENT POLICIES INTERIM ENFORCEMENT POLICIES Interim Interim Enforcement Generally Licensed Enforcement Policy for Generally Licensed Devices Containing Byproduct Containing Byproduct Material Material (10(10 CFR 31.5)

This section sets forth the interim enforcement enforcement policy that the NRC will follow to exercise enforcement enforcement discretion for certain violations of requirements requirements in 10 CFR Part 31 for generally licensed devices devices containing byproduct byproduct material. It addresses violations that persons licensed licensed pursuant pursuant to 10 CFR 31.5 identifY identify and correct now, as well as during the initial cycle of of the notice and response response program contemplated contemplated by the proposed new requirements requirements published published in the Federal Register Register on December 1998 (63 FR 66492), entitled "Requirements December 2, 1998 "Requirements for Those Those Who Possess Certain Industrial Devices Containing Byproduct MaterialMaterial to Provide Requested Requested Information".

Information" .

Exercise of Enforcement Enforcement Discretion Discretion Under this interim enforcement enforcement policy, enforcement enforcement action normally will not be taken for for violations of 10 CFR 31.5 if they are identified by the general general licensee, and reported reported to the NRC if reporting reporting is required, if the general licensee licensee takes appropriate corrective corrective action to address the specific violations and prevent recurrence specific recurrence of similar problems.

Exceptions Enforcement Enforcement action may be taken where there is: (a) (a) failure to take appropriate appropriate corrective corrective action to prevent recurrence of similar violations; (b) failure to respond and provide the information information required by the notice and response response program (if it becomes a final rule); (c) becomes afmal (c) failure to provide complete complete and accurate information to the NRC; or (d) a willful violation, such as willfully disposing of generally generally licensed material in an unauthorized unauthorized manner. Enforcement sanctions in these cases may include include civil penalties penalties as well as Orders to modify modifY or revoke the authority to possess radioactive sources under the general license.

t~e general 66

Interim Enforcement Enforcement Policy Regarding Enforcement Enforcement Discretion for Certain Fitness-for-Duty Issues (lO (10 CFR Part 26)

This section section sets forth the interim enforcement enforcement policy that the NRC NRC will follow to requirements in 10 CFR Part 26, Fitness-enforcement discretion for certain violations of requirements exercise enforcement for-Duty Programs that occur after December 30,2002. 30, 2002. The NRC will also exercise exercise enforcement discretion and normally not pursue past violations for insufficient suitable inquiries (where licensees failed to contact licensees contact employers when individuals individuals had worked employers for less than worked for employers 30 days) and past violations for failures to perform pre-access drug tests (where individuals were perform pre-access subject to a FFD program within the last 30 days) that occurred December 30,2002. The occurred prior to December subsequent Commission-approved policy, subject to subsequent Commission-approved associated policy, guidance, or regulation, is final revision of 10 CFR Part 26 is issued and becomes in effect until a [mal becomes effective.

Suitable Inquiry The regulation in 10 CFR 26.3 requires that before before granting an individual unescorted unescorted access, a licensee must conduct a suitable inquiry consisting inquiry consisting of a "best-effort verification verification of of employment history for the past five years, but in no case employment case less than three years, obtained obtained through employers to determine if a person contacts with previous employers contacts person was, in the past, tested positive for illegal drugs, subject to a plan for treating substance illegal substance abuse, removed from, or made ineligible for activities within the scope of 10 CFR Part 26, or denied unescorted access activities access at any other nuclear power plant or other employment power employment in accordance accordance with a fitness-for-duty policy."

The requirement does not provide individual is reinstated at a provide an exception when an individual licensee licensee facility or transferred transferred within a licensee corporation or to another licensee corporation another licensee licensee where there is little or no interruption in authorization. The term, "authorization,"

"authorization," refers to a period during which an individual individual maintained unescorted access or was assigned maintained unescorted assigned to perform activities activities within the scope of Part 26. However, enforcementenforcement action will not normally be taken for failure to contact employers, if the following practice is adopted:

contact interim employers, individual applicant's authorization If the individual calendar days or authorization has been interrupted for 30 calendar or less and the individual's individual's last authorization authorization was terminated terminated favorably, before granting before granting unescorted access to the protected authorization for unescorted protected area of a nuclear nuclear power plant or assigning the individual to perform activities within the scope scope of Part 26, the licensee shall obtain and verify verifY alcohol-related arrests) for the period since the (i.e., a report of any drug- or alcohol-related that a self-disclosure (i.e.,

authorization contains last authorization potentially disqualifying contains no potentially disqualifYing FFD information, unless the individual was subject licensee-approved behavioral subject to a licensee-approved behavioral observation arrest-reporting program observation and arrest-reporting throughout the period of interruption. Potentially disqualifYing FFD information Potentially disqualifying information means demonstrating that an individual has, during the period authorization was information demonstrating interrupted:

(1) Violated employer's drug and alcohol testing policy; Violated an employer's (2)

(2) Used, sold, or possessed illegal drugs; 67

(3)

(3) Abused legal drugs; (4) Subverted or attempted to subvert subvert a drug or alcohol testing program; (5) Refused Refused to take a drug or alcohol test; (6)

(6) Been subjected to a plan for substance substance abuse treatment (except (except for self-referral);

self-referral); or or (7)

(7) Had legal or employment employment action taken for alcohol alcohol or drug use.

The licensee shall also ensure that the individual has met FFD refresher training requirements.

The requirements requirements also do not provide an exception for each licensee to conduct a suitable inquiry inquiry into an individual applicant's past five years of employment when an individual individual is reinstated at a licensee facility or transferred to another licensee licensee facility. However, enforcement enforcement action will not normally be taken for failure to contact employers from the past five years, if the following practice is adopted:

Licensees Licensees may rely upon the information gatheredgathered by previous licensees licensees regarding regarding an individual applicant's past five years of employment employment to meet the suitable inquiry inquiry requirement.

requirement.

The The NRC NRC may take enforcement enforcement action when a licensee licensee does not follow these practices.

practices.

Pre-access Testing Pre-access Testing The regulation The regulation in 10 CFR 26.24(a)(1) 26.24( a)(1) requires that a person be tested for drugs and and alcohol "within 60 days prior to the initial granting granting of un unescorted escorted access to protected areas."areas."

The requirement The requirement does not provide an exception exception when an individual is reinstated reinstated at a licensee facility licensee facility or transferred within a licensee corporation corporation or to another licensee where there is little or no interruption interruption in authorization. However, enforcement enforcement action will not normally be taken for failure to conduct a pre-access test for alcohol and drugs, if the following practice practice is adopted:

If the individual applicant's applicant's authorization has been interrupted interrupted for 30 calendar days or or less and less and the individual's authorization was terminated favorably, in order to grant last authorization authorization authorization for unescorted unescorted access to the protected area of a nuclear power power plant or assigning the individual individual to perform activities within the scope of Part 26, the licensee shall:

(1) Obtain and verify verify that a self-disclosure for the past 30 days reveals no potentially disqualifying disqualifYing information, unless the individual was subject to a licensee-approved licensee-approved behavioral observation behavioral observation and arrest-reporting arrest-reporting program program throughout the period period ofof interruption; interruption; and (2)

(2) Ensure that Ensure that the the individual individual has met FFD refresher refresher training requirements.

68

If the individual applicant's authorization lfthe authorization has been interrupted for 31 days to 60 days and individual's last authorization the individua1's authorization was terminated favorably, in order to grant authorization for unescorted access to the protected area of a nuclear power plant or assigning the individual individual to perform activities within the scope of Part 26, the licensee shall:

(1) Obtain and verifY verify that a self-disclosure self-disclosure for the period since the last authorization authorization contains contains no potentially disqualifying FFD information, unless the individual was subject to a potentially disqualifYing licensee-approved licensee-approved behavioral behavioral observation observation and arrest-reporting arrest-reporting program throughout the period of interruption; (2)

(2) Within 5 working days of granting granting authorization, complete a suitable inquiry for the period since last authorization was terminated, terminated, unless the individual individual was subject to a licensee-approved behavioral licensee-approved behavioral observation observation and arrest-reporting arrest-reporting program program throughout the period of interruption; (3) Verify VerifY that results of an alcohol test are negative and collect a specimen for drug testing, unless either a drug and alcohol test meeting meeting the standards standards of Part 26 was performed performed within the past 60 days and results were were negative or the individual was subject to a licensee-approved Part 26 FFD program licensee-approved program that included random drug and alcohol testing throughout throughout the period of interruption; and (4) Ensure that the individual has met FFD refresher training requirements.

The NRC may take enforcement enforcement action when a licensee does not follow these practices.

69

Interim Enforcement Enforcement Policy Regarding Regarding Enforcement Enforcement Discretion for Certain Certain Fire Protection Protection Issues (10 (10 CFR 50.48)

This section sets forth the interim enforcement enforcement policy that the NRC will follow to exercise enforcement exercise enforcement discretion for certain certain violations of requirements requirements in 10 CFR 50.48, Fire protection (or fire protection fue protection license conditions) that are identified as a result of the transition to a new risk-informed, performance-based new performance-based fIre fire protection protection approach included in paragraph (c) (c) of of 10 CFR 50.48 and for certain existing identified noncompliances existing identifIed noncompliances that reasonably may be resolved by compliance by compliance with 10 CFR 50.48(c). Paragraph (c) (c) allows reactor licensees licensees to voluntarily comply with the risk-informed, performance-based fire risk-informed, performance-based fue protection approaches approaches in National Fire Protection Association Protection Association (NFPA) Standard Standard 805 (NFPA 805), "Performance-Based "Performance-Based Standard Standard For For Fire Protection Fire Protection For Light Water Reactor Electric Generating Plants,"

Water Reactor Electric Generating Plants," 2001 Edition (with limited exceptions stated in the rule language).

exceptions language).

For those noncompliances identified during the licensee's transition noncompliances identifIed transition process, this enforcement discretion policy will be in effect for up to two years from the date of a licensee's enforcement licensee's letter of intent to adopt the requirements requirements in 10 CFR 50.48(c) and will continue to be in place until NRC approval of of the license amendment request request to transition to 10 CFR 50.48(c).50.48(c). This discretion discretion policy may be extended upon a request from the licensee with adequate adequate justification.

If, If, after SUbmitting submitting the letter of intent to comply with 10 CFR 50.48(c) 50.48( c) and before submitting the license amendment request, the licensee determines submitting determines not to complete complete the transition transition to 10 CFR 50.48(c), the licensee must submit a letter stating their intent to retain their existing license basis and withdrawing license withdrawing their letter of intent to comply with 50.48(c). 50.48( c). Any violations identified identifIed prior to the date of the above withdrawal withdrawal letter will be eligible eligible for discretion, provided they are resolved resolved under the existing licensing basis and meet the criteria included in this policy policy for these violations. Violations identified after the date of the above withdrawal Violations identifIed withdrawal letter will be dispositioned in in accordance accordance with normal enforcement enforcement practices.

A. Noncompliances Noncompliances IdentifiedIdentified During the Licensee's Licensee's Transition Process Under this interim enforcement enforcement policy, enforcement action normally will not be taken for aa violation violation ofof 10 10 CFR 550.48(b) 0.48(b) (or the requirements in a fue fire protection protection license condition) involving aa problem involving problem such such as in engineering, engineering, design, implementing implementing procedures, procedures, or installation, ifif the violation is documented in an inspection inspection report and it meets all of the following criteria:

(1)

(1) It It was was licensee-identified licensee-identifIed as as aa result of its result of its voluntary initiative initiative to adopt the risk-informed, performance-based fire protection performance-based fue protection program program included under 10 CFR 50.48(c) or, if the NRC identifies identifies the violation, it was likely in the NRC staff's staffs view that the licensee would have identifIed identified the violation in light of the defined scope, thoroughness,thoroughness, and schedule of of the licensee's transition transition to 10 CFR 50.48(c) 50.48(c) provided provided the schedule reasonably providesprovides forfor completion of the transition within two years of the date of the licensee's letter of intent to to implement 10 CFR 50.48( 50.48(c) c) or other period granted by NRC; 70 70

(2)

(2) It was corrected corrected or will be corrected as a result of completing the transition to 10 CFR 50.48(c).

50.48( c). Also, immediate immediate corrective corrective action and/or compensatory measures are compensatory measures within a reasonable taken within reasonable time commensurate significance of the issue commensurate with the risk significance identification (this action should involve expanding following identification expanding the initiative, as necessary, identify other issues caused by similar root causes);

to identifY (3)

(3) It was not likely to have been previously identified by routine licensee licensee efforts such as normal surveillance surveillance or quality assurance assurance (QA) activities; and (4) It was not willful.

The NRC may take enforcement enforcement action when these conditions are not met or when a violation violation that is associated associated with a finding of high safety significance significance is identified.

While While the NRC may exercise exercise discretion for violations meeting the required required criteria criteria where the licensee licensee failed to make a required report to the NRC, a separate separate enforcement action will normally be issued for the licensee's licensee's failure to make a required report.

B. Existing Identified Noncompliances In addition, licensees may have have existing identified noncompliances noncompliances that could reasonably reasonably be corrected corrected under 10 CFR 50.48(c). For these noncompliances, noncompliances, the NRC is providing enforcement discretion for the implementation enforcement implementation of corrective corrective actions until the licensee licensee has transitioned to 10 CFR 50.48( 50.48(c)c) provided provided that the noncompliances noncompliances meet all of the following criteria:

(1) The licensee has entered the noncompliance noncompliance into their corrective corrective action program and implemented appropriate implemented appropriate compensatory compensatory measures, (2)

(2) associated with a finding that the Reactor Oversight Process The noncompliance is not associated Process Significance Determination Process Significance Determination Process would evaluate evaluate as Red, or it would not be categorized categorized at Severity Level I, I,

(3)

(3) The licensee submits a letter of intent by December December 31, 31, 2005, 2005, stating its intent to transition to 10 CFR 50.48(50.48(c).

c).

After December December 31, 31, 2005, 2005, as addressed in (3) (3) above, this enforcement enforcement discretion for implementation corrective actions for existing identified noncompliances implementation of corrective noncompliances will not be available and the requirements requirements of 10 CFR 50.48(b) (and any other requirementsrequirements in fIfe fire protection license license conditions) will be enforced enforced in accordance accordance with normal enforcement enforcement practices.

71

Enforcement Policy Regarding the Use of Alternative Dispute Interim Enforcement Resolution Dispute Resolution l.

I. Introduction Introduction A. Background

Background

This section sets forth the interim enforcement policy that the NRC will follow to interim enforcement undertake undertake a pilot program testing the use of Alternative Alternative Dispute Resolution Resolution (ADR) in the enforcement enforcement program.

B. Scope Scope The pilot program program scope scope consists of the trial use of ADR for cases involving: (1) alleged discrimination for engaging engaging in protected protected activity prior to an NRC investigation; investigation; and (2)

(2) both discrimination discrimination and other wrongdoing wrongdoing cases after the Office of Investigations competed an investigation. Specific points in the Investigations has competed enforcement process where ADR may be requested are specified below.

enforcement Mediation will be the form of ADR typically utilized. Certain Certain cases may only require facilitation, a process process where the neutral's function is primarily to support communication process rather than focusing on the parties reaching a the communication settlement.

Note: Although Although the NRC's ADR program may cause the parties to negotiate issues which may also form the basis for a claim under Section Section 211 of the Energy Reorganization Reorganization Act of 1974, 1974, as amended, the Department of Labor's (DOL) timeliness requirements requirements for filing a claim are in no way altered by the NRC's program.

In cases involving an allegation allegation of discrimination, any underlying underlying technical issue will be treated treated as a separate separate issue, or concern, within the allegation allegation program. The allegation program will be used to resolve concerns (typically (typically safety concerns) concerns) and issues other than the discrimination complaint.

II.

H. General A. Responsibilities Responsibilities and Program Administration Administration The Director, OE, is responsible for the overall overall program. In addition, the Director, OE, will serve as the lead NRC negotiator negotiator for cases cases involving discrimination discrimination after 01 completes an investigation. The Director, OE, may also also designate the Deputy Deputy Director, OE, to act as the lead negotiator.

72 72

Regional Administrators Regional Administrators are designated designated as the lead NRC negotiator negotiator for cases wrongdoing other than discrimination. The Regional Administrator involving wrongdoing Administrator may designate the Deputy Deputy Regional Administrator to act as the lead negotiator Regional Administrator negotiator oror the Director or Deputy Director, OE, may also serve as the lead negotiator negotiator for other wrongdoing cases.

The Program Administrator Administrator will provide programprogram oversight oversight and support for each region and headquarters headquarters program offices. Program and neutral evaluations will be provided to the Program Program Administrator. The Program AdministratorAdministrator may serve as the intake neutral for post investigation ADR. An "intake "intake neutral" develops information information and processes information for mediation. As an intake neutral, the confidentiality confidentiality provisions provisions discussed below will apply.

The Office Allegation Allegation Coordinators Coordinators (OACs) are normally a complainant's first frrst substantive substantive contact contact when a concern regarding discrimination is raised. As such, regarding discrimination the OACs may serve as an intake neutral develop~ information and processes neutral who develops processes the necessary information for mediation mediation under Early ADR. The OAC has the option to refer the whistleblower whistleblower to the third party neutral to process the necessary information information for mediation under Early ADR. The confidentiality confidentiality provisions in Section II.B.7 will apply to the OAC, third party intake neutral, and Program Section 11.B.7 Program Administrator. The OAC will also process documentation documentation necessary necessary to operate operate the program.

B. General Rules/Principles General Rules/Principles Unless specifically addressed in a subsequent subsequent section, the rules described in this section section apply generally throughout the ADR program, regardless of where where in the overall enforcement process the ADR sessions occur.

overall enforcement 1.

I. Voluntary. Use of the NRC ADR program is voluntary, and any Voluntary.

participant participant may end the mediation at any time. The goal is to obtain an an agreement satisfactory agreement satisfactory to all participants on issues in controversy.

2. Neutral qualification. Generally, Neutral qual[fication. Generally, a neutral should be knowledgeable knowledgeable and experienced experienced with nuclear nuclear matters matters or labor and employment employment law. However, any neutral neutral that is satisfactory satisfactory to the parties parties acceptable.

is acceptable.

3. neutrals. OE will maintain a list of organizations Roster of neutrals. organizations from which services of neutrals neutrals could be obtained. The parties parties may select a mediator mediator from any of these organizations; organizations; however, the parties are not required to use the organizations provided provided and any neutral mutually agreeable agreeable to the parties acceptable.

parties is acceptable.

73

4. Mediator selection. If the parties have not selected Mediator selection. selected a mediator within fourteen days, the Program Administrator Administrator or OAC may propose propose a mediator mediator for the parties' consideration.

5.

5. Neutrality. Mediators Neutrality. Mediators are neutral. The role oftheof the mediator is to provide an environment where all participants participants will have an opportunity opportunity to resolve their differences. The parties should each consult consult an attorney or other professional professional if any question of law, content of a proposed agreement agreement on on issues in controversy, or other issues exists.

For Early ADR, the OAC or third party neutral will serve as an intake neutral. Should any party seek to discuss the NRC's enforcement enforcement ADR process in detail, the party should be referred referred to the OAC or third party neutral. The OAC will initiate discussion of the option to mediate and process the necessary documentation. Subsequently, for post investigation investigation ADR, the program administrator or third party neutral will serve as the intake neutral. Due to the nature of conversations conversations that typically occur occur between an intake neutral and the parties, these conversations will also be considered considered confidential.

6. Mediation sessions. Once selected Mediation sessions. selected by the parties and contracted contracted by the OAC or third party intake neutral, the mediator mediator will promptly contact each of the parties parties to discuss the mediation process under the Program, reconfirm party interest in proceeding, establish a date and location for the mediation session and obtain any other information s/he believes likely to be useful. The mediator mediator will preside over all mediation sessions, and will be expected to complete the mediation within 90 days after referral unless the parties, and the NRC if not a party, agree otherwise. At the conclusion conclusion of the mediation, parties parties will be asked to fill out and submit an evaluation evaluation form for the mediator mediator that will be sent to the Program Administrator.

Normally, a settlement settlement is expected to be reached reached and signed within 90 days from when the parties parties agree to attempt ADR. A principal reason for for Early ADR is the quick resolution of the claim, thereby improving the environment (SCWE). If the parties cannot agree to safety conscious work environment a settlement within 90 days, the NRC must assume a settlement will not be reached and continue continue with the investigation enforcement process.

investigation and enforcement Where good cause is shown and all parties agree, the NRC may allow a small extension extension to the 90 day limit to allow for completion completion of a settlement agreement.

Settlement agreements Settlement agreements in Early ADR will not be final until 3 days after the agreement has been signed. EitherEither party may reconsider reconsider the settlement 74 74

agreement during the 3 day period. Subsequent Subsequent concerns concerns regarding implementation of the settlement implementation agreement should be directed to the settlement agreement neutral, or if necessary, the OAC.

7. Confidentiality. The mediator will specifically Confidentiality. specifically inform all parties and other attendees that all mediation activities activities under the Program Program are subject to the confidentiality provisions of the Administrative provisions ofthe Administrative Dispute Resolution Act, 5 U.S.C. 574; the Federal ADR Council's guidance document document entitled "Confidentiality in Federal ADR Programs;"

"Confidentiality Programs;" and the explicit confidentiality terms set forth in the Agreement confidentiality Agreement to Begin Begin Voluntary Mediation signed by the parties. The mediator will explain explain these confidentiality terms and offer to answer questions regarding them.

confidentiality 8.

8. Good Faith.

Faith. All participants participants will participate in good faith in the mediation mediation process process and explore potentially feasible options that could lead to the management or resolution of issues in controversy.

9. Not legal legal representation.

representation. A mediator is not a legal representative representative or legal counsel. The mediator will not represent represent any party in the instant case or or any future proceeding proceeding or matter relating to the issues in controversy controversy in this case. The mediator is not either party's lawyer and no party should rely on the mediator mediator for legal advice.

10.

lO. MediatorFees.

Mediator Fees. If Early ADR (defined below) is utilized, the NRC, subject to the availability availability of funds, will pay the mediator's entire fee. For For cases where a licensee requests ADR subsequent subsequent to the completion of an 01 report, the licensee requesting requesting ADR will pay half of the mediator's mediator's fee fee and the NRC, subject to the availability of funds, will pay half. half. The NRC will recover recover the mediator mediator fees it pays through annual fees assessed to licensees under 10 CFR Part 171. 171.

11.

11. Exceptions. The only exception Exceptions. exception to the offering of Early ADR by the NRC will be abuse ofthe of the program, e.g.,

e.g., a large number of repetitive requests for ADR by a particular facility, contractor, or whistleblower. Should the NRC believe the ADR program has been abused in some manner by one of of the parties potentially involved, the Director, OE will be notified.

To maximize the potential potential use of the ADR pilot program, for cases after after an 01 investigation investigation is completed, the NRC will Will at least consider consider negotiating a settlement settlement with a licensee for any wrongdoing case if requested. However, there may be certain certain circumstances circumstances where it may not not be appropriate for the NRC to engage in ADR.

75

12. Number of settlement attempts.

attempts. Each case will be afforded afforded a maximum maximum of of underlying two attempts to reach a settlement on the same underlying issue through the use of ADR. An "attempt" "attempt" is defined as one or more mediatedmediated sessions conducted conducted at a specific specific point in the NRC's enforcement process enforcement process (generally (generally within a 90 day period). However, in general, settlement at any time without the use ofaof a neutral is not precluded by the ADR program.

13.

13. Finality. Cases that reach a settlement Finality. settlement (and are acceptable acceptable to the NRC),

either either in Early ADR or after an 01 investigation investigation is complete, constitute constitute a enforcement decision on the case by the NRC.

final enforcement .

Ill.

111. Opportunities ADR Opportunities A. Licensee Sponsored Programs Licensees Licensees are encouraged encouraged to develop ADR programs of their own for use in programs oftheir conjunction with an employee concerns type program. If an employee employee who engaging in protected alleges retaliation for engaging protected activity utilizes a licensee's program to settle the discrimination discrimination concern, either either before or after contacting contacting the NRC, NRC, the licensee may voluntarily report the settlement to the NRC as a settlementsettlement within the NRC's jurisdiction. If notified of the settlement settlement prior to initiation of an investigation, the NRC will review the settlement for restrictive agreements restrictive agreements potentially in violation violation of 10 10 CFR 50.7(f),

50.7(f), or other, similar regulations. Assuming Assuming no such restrictive restrictive agreements exist, the NRC will not investigate investigate or take enforcement action.

enforcement B. Early ADR The term "Early ADR"ADR" refers to the use of ADR prior to an 01 investigation. The parties parties to Early ADR will normally normally be the complainant complainant and the licensee. If the complainant is an employee of a licensee contractor, the parties complainant parties will be the complainant and the contractor. Generally, the Early ADR process will parallel parallel and work in conjunction with the NRC allegation allegation program.

The allegation allegation process process will be used through the determination determination of a prima facie case. If an Allegation Allegation Review Board (ARB) determines determines a prima facie case exists, the ARB will normally normafly recommend recommend the parties be offered the opportunity opportunity to use Early ADR. Exceptions Exceptions to such a recommendation should be rare and be based solely on an identified identified and articulated abuse of the ADR process by a party who would be involved in the case under under consideration. Exceptions Exceptions will be approved by the Director, OE, prior to initiating an investigation investigation based based on denial of ADR.

76 76

Early ADR cases will be tracked in the Allegation Management Management System (AMS).

However, the allegation process timeliness timeliness measurement measurement will be stayed onceonce the ARB determines determines that ADR should be offered in time ADR is offered until the point in.time declined by either party or the case is settled.

agreement is reached, the mediator will record the terms ofthat When an agreement of that agreement. The parties may sign the agreement agreement at the mediation session, or any party may review the agreement agreement with his/her his/her attorney before the document is placed placed in final form and signed. However, as noted above, settlement agreementsagreements in Early ADR will not be final until at least 3 days after the agreement has been been signed. No participant participant will hold the NRC liable for the results of the mediation, whether whether or not a resolution is reached.

A settlement settlement agreement agreement between the parties will be reviewed by the NRC. OE will coordinate coordinate the review with the Office of the General Counsel Counsel (OGC). The review review will ensure ensure that no restrictive agreements agreements in violation of 1100 CFR 50.7(f) or or other NRC regulations regulations are contained in the settlement settlement and will normally normally be completed within 5 working days of receipt. Given an acceptable settlement, settlement, the NRC will not investigate or take enforcement enforcement action.

The NRC expects that parties to Early ADR will agree to some form foim of of confidentiality. However, However, that agreement agreement cannot extend to the reporting of any safety concerns potentially potentially discussed discussed during the ADR sessions sessions if one of the parties desires to report the concern. Either party may report safety concerns concerns discussed during ADR sessions to the NRC without regard regard to confidentiality confidentiality agreements.

agreements.

Safety concerns and their disposition disposition may be discussed between between the parties ifif desired. In cases where an Early ADR negotiation is between between a licensee contractor and the contractor's contractor's employee, employee, the NRC expects the contractor contractor to ensure the licensee is aware of any safety issues discussed during the negotiations.

In addition to the settlement settlement agreement, agreement, the licensee should provide the NRC with any planned planned or completed actions relevant relevant to the safety safety conscious conscious work environment that the licensee licensee has determined determined to be appropriate.

Generally no press release or other public announcement will be made by the NRC for cases settled by early early ADR. However, all documents, documents, including including the proposed proposed settlement settlement agreement, agreement, submitted to the NRC will be official agency records, and while not generally generally publicly available, still subject to the Freedom of of Information Act (FOIA).

associated with processing Documents associated processing an Early ADR case will not generally be publicly publicly available, available, consistent with the allegation allegation program. However, documents 77 77

may be subject to the FOIA and may be released, released, subject to redaction, pursuant to a FOIA FOJA request.

Some negotiations may fail to settle the case. When a settlement is not reached, intake neutral will be notified, typically by the mediator, and an the appropriate intake ARB will determine the appropriate accordance with the allegation.

appropriate action in accordance allegation.

program.

C. Post-Investigation ADR Post-Investigation ADR Post-investigation ADR refers to the use of ADR anytime after an OJ Post-investigation 01 investigation investigation complete and an enforcement is complete concludes that pursuit of an enforcement enforcement panel concludes enforcement action appears warranted. Generally, post- investigation ADR processes processes will conjunction with the NRC enforcement parallel and work in conjunction parallel enforcement program.

generally three issues that can be After an investigation is complete, there are generally violation occurred, the appropriate enforcement resolved using ADR; whether a violation resolved occurred, enforcement appropriate corrective action, and the appropriate vio lation( s). If the parties corrective actions for the violation(s).

considered in an ADR session.

agree, any or all three may be considered Two different types of enforcement enforcement cases will be eligible for ADR after an investigation is complete, discrimination investigation wrongdoing cases. ADR will discrimination and other wrongdoing enforcement process after 01 has considered at three places in the enforcement normally be considered completed an investigation: (1) (1) After an enforcement concluded there is enforcement panel has concluded the need to continue pursuing enforcement action based on an 01 case pursuing potential enforcement enforcement conference and prior to the conduct of a predecisional enforcement conference (PEC); (2) (2) enforcement action is taken, typically a Notice of Violation (NOV) after the initial enforcement and potentially a proposed civil penalty; and (3) after imposition of a civil penaltypenalty and prior to a hearing request.

session after an 01 investigation is complete will be the The parties to an ADR session licensee and the NRC. Fees associated with the neutral will typically be divided divided between the NRC and the licensee, with each paying half of between the total cost.

ofthe expected to be complete Settlement discussions are expected Settlement complete within 90 days of initiating ADR prior to a PEC. The NRC may withdraw withdraw from settlement discussions if negotiations have not been completed completed in a timely manner.

agreement will normally be confirmed by order.

The terms of a settlement agreement settlement will be agreed to during the negotiation.

Typically, the specific terms of settlement The staff will then incorporate incorporate appropriate confirmatory order, a draft appropriate terms into a confirmatory of which will then be agreed to by the licensee prior to issuance.

78 78

If an attempt to resolve a case using ADR prior to the conduct conduct of a PEC fails, a predecisional predecisional enforcement conference will normally be offered enforcement conference offered to the licensee.

The PEC will be conducted conducted as described in the Enforcement Enforcement Policy.

For cases within the scopescope of the pilot program, after a panel concludes concludes that a case warrants continuation of the enforcement warrants continuation enforcement process, process, the responsible region region or office will contact the licensee licensee and offer either a PEC or ADR. Consistent Consistent with the Enforcement Enforcement Policy, a written written response could could be offered offered at the staffs discretion.

Public notification notification of the settlement settlement will normally normally be a press release and the confirmatory order confirmatory order will be published in the Federal FederalRegister.

Register.

Confidentiality with the NRC as a party will be determined determined by the parties as allowed by the ADR Act.

1.

1. Discrimination Discrimination Cases Consistent with centralization centralization of ofthe discrimination enforcement the discrimination enforcement process, the Director, Enforcement, will normally Director, Office of Enforcement, normally negotiate for the NRC.

Normally Normally the NRC will coordinate participation of the complainant.

coordinate participation complainant.

While the complainant will not be a party to the ADR process after 01 investigation report, the NRC issues an investigation NRC will typically seek the complainant's input to the process. Normally, the NRC will at least seek input firom from the complainant regarding suggested corrective corrective actions aimed aimed at improving the safety conscious work environment.

environment.

01 reports (not including exhibits) will normally normally be provided provided to the licensee when the choice of ADR or a PEC is offered.

A licensee licensee may request request ADR for discrimination violations based solely on on a finding by DOL. However, the staff will not negotiatenegotiate the finding by DOL. The appropriate appropriate enforcement enforcement sanction and corrective corrective actions will be the typical focus of settlement settlement discussions.

2. Other Than Discrimination Discrimination Wrongdoing Wrongdoing The regional administrator will normally be the principal negotiator for the regional administrator NRC in ADR sessions on other wrongdoing wrongdoing cases. AfterAfter imposition of a civil penalty or other order, the Director, Director, Office of Enforcement Enforcement and applicable regional administrator applicable administrator may determine determine that the Director Director would would be the appropriate appropriate negotiator.

79 79

Typically, an enforcement enforcement panel will be conducted to discuss the NRC's NRC's specific interests specific interests in the case prior to the regional administrator attending the settlement discussions. A limited review of the settlement terms may be conducted conducted in conjunction with the preparation preparation of the confirmatory confirmatory order.

The 01 or report will not routinely be offered to the licensee prior to ADR.

However, the or 01 report may be provided, as necessary, necessary, during the negotiations with the licensee.

IV. Integration With Traditional Enforcement Enforcement Policy A. Potential Potential Future Enforcement Enforcement Actions Civil Penalty Penalty AssessmentsSection VI.C.2 VLC.2 of the Enforcement Enforcement Policy provides provides the method for determination determination of a civil penalty amount. One aspectaspect of the determination determination uses enforcement history as a factor. If the staff considers a civil penalty rfthe penalty for a future escalated enforcement action, settlements enforcement settlements under enforcement ADR program occurring under the enforcement occurring after a formal enforcement action is taken (e.g. an NOV is issued) may count as an enforcement action an enforcement enforcement case for purposes of determining determining whether identification identification credit is considered. Settlements occurring occurring prior to an or 01 investigation investigation will not count as previous enforcement.

enforcement. The status of settlement settlement agreements agreements occurring occurring after anan investigation investigation is completed but prior to an NOV being issued issued will be established established as part of the negotiation between the parties.

80

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OFFICE OF SECRETARY RULEMAKINGS RULEMAKINGS AND ADJUDICATIONS STAFF ADJUDICATIONS OMB Control No.: 3150-0012 3150-0012 UNITED UNITED STATES NUCLEAR REGULATORY

.NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION OF.NUCLEAR WASHINGTON, D.C. 20555:0001 WASHINGTON, August 3, 2001 NRC BULLETIN 2001-01:

2001-01: CIRCUMFERENTIAL CRACKING CIRCUMFERENTIAL CRACKING OF REACTOR PRESSURE VESSEL HEAD PENETRATION NOZZLES HEAD PENETRATION NOZZLES Addressees All holders of operating licenses licenses for pressurized pressurized water nuclear nuclear power reactors, except except those who have ceased operations and have certified certified that fuel has been been permanently permanently removed removed from thethe reactor reactor vessel.

Puroose Purpose U.S. Nuclear The U.S. Regulatory Commission (NRC) is issuing this bulletin Nuclear Regulatory bulletin to:

(1) request that addressees addressees provide information related to the structural provide information structural integrity of the reactor pressure penetration (VHP) nozzles pressure vessel head penetration nozzles for their respective respective facilities, including the extent of VHP nozzle ofVHP nozzle leakagEtand leakage~and crackingcracking that has been found to date, the the inspections and repairs that have undertaken to satisfy applicable have been undertaken applicable regulatory requirements, and the basis for concluding that their plans requirements. plans for future inspections will ensure compliance with ensure With applicable requirements. and applicable regulatory requirements, and (2)

(2) addressees provide to require that all addressees to the NRCNRC a written response response in accordance accordance with the the provisions of 10 10 CFR 50.54{f}.

50.54(f).

Backaround

Background

The recent recent discoveries discoveries of cracked cracked and leaking Alloy 600 VHP nozzles, including control rod drive drive mechanism (CRDM) mechanism (CRDM) and thermocouple nozzles, at four pressurized pressurized water reactors (PWRs) concerns about the structural integrity of have raised concems VHP nozzles throughout ofVHP throughout the PWR industry.

Nozzle cracking at Oconee Nuclear Nozzle craCking Nuclear Station Unit 1 (ONS1) (ONS1) in November 2000 2000 and Arkansas Arkansas Nuclear Unit 1I (ANO1)

Nuclear One Unit (AN01) in February 2001 was limited to axial cracking, an occurrence occurrence deemed deemed to be of limited safety concern in the NRC staffs generic safety evaluation limited safetyconcem evaluation on the the cracking ofVHPnozzles.

of VHP nozzles, dated November 19, dated November 1993. However.

19.1993. However, the discovery discovery of circumferential cracking at Oconee Nuclear Station Unit Oconee Nuclear Unit 3 (ONS3) in FebruaryFebruary 2001 and Oconee Nuclear Nuclear Station Unit 2 (ONS2) in April 2001 S particularly particularly the largelarge circumferential circumferential cracking identified in in two CRDM nozzles nozzles at ONS3 S S has raised concerns concemsabout about the potential safety implications implications and prevalence of cracking in VHP nozzles in PWRs.

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Page 2 of 15 Page20f described in NRC As described NRC Information Notice (IN)2001-05, "Through-Wall Circumferential Notice (IN) Cracking Ci(cumferential C(8cking of Reactor Pressure Vessel Head Control Control- Rod Drive MechanismMechanism Penetration Penetration Nozzles Nozzles at Oconee Nuclear OConee Nuclear Station, Unit.3,"-dated April 30, 2001, Station,Unit3,*-dated 2001, Duke Energy Corporation (the licensee)

Corporation (the performed aa visual

- perfonned examination (VT-2) on the outer surface of the reactor pressure visual examination pressure vessel vessel

• (RPV)

(RPV) head ONS3 to inspect for indications of borated water leakage, as part of normal head at ONS3 surveillance during a planned maintenance surveillance maintenance outage. This Visual visual examination followed cleaning of the RPV head during the prior outage outage to remove all existing boric acid deposits (from other sources such as leaking CRDM sources CROM flanges) that could mask the identification identification of subsequent subsequent indicative of new deposits that would be indicative new or ongoing leakage. The VT-2 examination revealed small amounts of boric acid deposits (less than 1 cubic CUbic inch) at locations where where the CRDMCROM nozzles exit the RPV head for nozzles far 9 of the 69 CRDM CROM nozzles. Subsequent Subsequent nondestructive nondestructive (NDE) identified 47 recoi'dable examination (NOE) recordable crack crack indications in in the 9 degraded degraded CRDM CROM nozzles.

The licensee initially characterized characterized these flaws as being axial and aa part.of part of the RPV pressure pressure boundary, or below-the-weld below-the-weld circumferential indications (which are not part of the RPV pressure circumferential indications pressure boundary), and initiated initiated repairs of the degraded areas.

Subsequent dye-penetrant Slibsequent dye-penetrant testing (PT) of the repaired areas revealed the presence presence of additional in two of the nine degraded indications in degraded nozzles. While repairing the indications indications inin these two nozzles, the licensee found that each nozzle nozzle had a circumferential extended about circumferential crack that extended 1650 1650 around the nozzle, above the weld (Le., (i.e., at aLaa location location that is part of theRPV the RPV pressure pressure boundary).

boundary). Further investigation investigation and.metallurgical and metallurgical examination examination identified identified that these cracks had initiated from the outside diameter initiated diameter (00)

(OD) ofthe of the CROM Thecircumferential crack in CRbM nozzles. The-circumferential in one one of the nozzles was through-wall, through-wall, and the crack crack in the other nozzle had pin hole indications indications on the the nozzle inside diameter diameter (ID).

(10). These cracks followed the contour of the weld profile.

The licensee stated that pre-repair ultrasonic testing (UT) examinations pre-repair Ufirasonic examinations had identified identified indications indications in these areas, but that these indications indications had been misinterpreted misinterpreted as inconsequential inconsequential craze craze cracking with unusual unusual characteristics.

characteristics. The characterizations characterizations of these two nozzle indications indications subsequently revised following the initial post-repairPT were subsequently post-repair PT examinations. The -licenseelicensee concluded that the root cause of the CRDM CROM nozzle cracking cracking was primary water stress corrosioncorrosion cracking (PWSCC), The cracking cracking initiated initiated at the OD 00 of the nozzles nozzles after cracking of the J- J-groove weld (see below) or adjacent heat-affected zone metal adjacent heat-affected metal permitted permitted coolant leakage leakage into the the

.annular annular region between between the CROM CRDM nozzle and the RPV head. This conclusion was based on metallurgical examinations, crack location metallurgical location and orientation, and finite elementanalyses.

element analyses.

The CRDM nozzles at ONS3 are approximately approximately 5 feet long and areJ-groove are J-groove welded welded to the inner radius of the RPV head, with the lower lower end of each nozzle extending extending about 6 inches below the the inside of the RPV head (see Attachment). The nozzles are constructed from4-inch from 4-inch OD 00 Alloy 600 Inconel procured 600lnconel procured in in accordance accordance with the requirements Specification SB-167 to the 1965 requirements of Specification 1965 Edition, including Addenda through the Summer 1967 Addenda,of Addenda, of Section Section 1I0fthe IIof the American Society Society of of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code. The weld Mechanical Engineers preparation preparation for the installation for the installation of each each nozzle in in the RPVhead RPV head was accomplished accomplished by BL 2001-01 Page Page 3 3 of 15 S 14D-04665 S140-04665 NRC007 -2016 NRC007-2016

machining and buttering buttering the J-groove J-groove with Alloy 182 weld metal. The RPV head head was subsequently subsequently stress relieved and then the final machining of the CROM CRDM penetrations, including including the counternore, was accomplished. Each nozzle was then machined the counterbore, machined to final dimensions dimensions to appropriate design interference assure the appropriate interference fit between the RPVhead RPV head borebore and the 00 OD of the nozzle. The interference interference fit of the CROM CRDM nozzles was made using a shrink to install shrink fit process .to the CROM CRDM nozzles. In this process, the nozzles were cooled to at least -140O -140°F;F; they were .then then closure head penetration, and the entire assembly was allowed to warm to inserted into the dosure room temperature (700F (70°F minimum).

minimum). The CROM CRDM nozzles were tack welded and then permanently welded to the closure head using Alloy 182 permanently welded 182 weld metal. The manual manual shielded metal arc welding (SMAW)

(SMAW) process was used for both the tack weld and the J-groove weld. During During weld buildup, the weld was ground ground and PT PT inspected inch of the weld. The final inspected at each 9/32 inch weld surface was ground and PT inspected.

The design and fabrication fabrication process for the VHPs in all PWR plants is similar to that described forONS3.

for ONS3.

Since the issuance of NRC IN circumferential cracking IN 2001-05, circumferential cracking was identified in another CRDM CROM nozzle, atONS2. examination of the RPV head, Duke Energy Corporation During aa visual examination at ONS2. During Corporation identified boric .acid identified acid deposits in the vicinity of four CROM nozzles at ONS2. Subsequent CRDM nozzles Subsequent UT UT examination identified a single CROM examination identified CRDM nozzle with one aD-initiated OD-initiated circumferential craCk, having circumferential crack, having a crack depth of 0.070 inch ((-11% -11 % through-wall) through-wall) and a length of 1.26 inches inches (-

(-10%

10% of the the circumference).

circumference).

Cracking due to PWSCC Cracking PWSCC in PWR PWR CROMCRDM nozzles and other VHP nozzles fabricated from Alloy Alloy 600 is not a new issue:

issue; axial cracking in in the CRDrJi CRDM nozzles has been identified since the late late 19805.

1980s. In addition, numerous small-bore Alloy 600 nozzles and pressurizer heater numerous small-bore heater sleeves havehave experienced PWSCC. Generally, attributable to PWSCC.

experienced leaks attributable components are exposed to high Generally, these components high temperatures (greater than 550° temperatures F).and a primary water environment. However, circumferential 5500 F)and cracking from the nozzle nozzle OD00 to the 10,ID, above the weld, and cracking of the J-groove J-groove weld havehave been previously identified in PWRs.

not been As described described in Generic Generic Letter (GL) 97-01,97-01, "Degradation of Control Rod Drive Mechanism Mechanism Nozzle and OtherVessel Other Vessel Closure Head Head Penetrations," dated April 1, 1997, an action plan plan was was implemented by the NRC staff in 1991 to address PWSCC of Alloy 600 VHP nozzles nozzles at all operating U.S. PWRs. After reviewing safety operating safety assessments submitted by the industryindustry and examining overseas overseas inspection findings, the NRC generic safety NRC staff concluded in its generic evaluation safety evaluation that CRDM nozzle nozzle and weld cracking cracking in PWRs was not an immediateimmediate safety concern.

concern. TheThe if PWSCC occurred (1) the cracks basis for this conclusion was that ifPWSCCoccurred cracks would be predominately predominately orientation, (2) the axial cracks would result in detectable leakage axial in orientation, leakage before before catastrophic catastrophiC failure (with the expectation expectation that CRDM nozzle thatCRDM cracking would result in a substantial nozzle cracking substantial volume of expected large amount of leakage would be detected leaking coolant) and (3) the expected during visual detected during surveillance walkdown performed as part of surveillance examinations performed walkdown inspections before damage before significant damage to the RPV head occurred. The safety evaluation identified concerns evaluation Identified concems about potential circumferential cracking (which would need to be addressed on a plant-specific circumferential plant-specific 5140-04666 46 66 S14D-0 NRC007-2017 NRC007-2017

BL2001-01 BL 2001-01 Page Page40f4 of 15 15 as aa consequence basis) as consequence of high high residual residual stresses resulting from initial stresses resulting manufacture and the initial manufacture the impact of tube impact tube straightening straightening that may have b~n needed have been after welding.

needed after welding. The The safety evaluation evaluation also noted the need enhanced leakage need for enhanced leakage monitoring.

monitoring.

The generic responses of generic responses of licensees to GL 97-01 were predicated on were predicated on the development development of susceptibility ranking susceptibility ranking models models to to relate the the operating conditions conditions (in (in particular particular the operating operating temperature and temperature and time) for each plantplant to the plant's plant's relative susceptibility PWSCC. The susceptibility to PWSCC. The generic responses generic responses committed committed to surface examinations of surface examinations of the VHP nozzles at the plants VHP nozzles plants identified as having identified having the the highest highest relative relative susceptibility C.onsistent with the expectations susceptibility ranking. Consistent expectations expressed by the expressed the NRC NRC staff in GL 97-01,97-01. the surface examinations conducted surface examinations prior to November cOnducted prior November identified only limited 2000 Identified circumferential cracking below and circumferential cracking, and limited axial cracking. below the weld weld in the the base metal of CRDM nozzles. but CRDM nozzles, circumferential cracking above but no circumferential above the nozzle welds and no nozzle welds no cracking cracking in the Alloy 182 182 welds.

Discussion The recent identification of circumferential The CRDM nozzles at ONS2 and QNS3.

cracking in CRDMnozzies circumferential cracking ONS3, along along with axial cracking in the J-groove J-groove welds at these these two ONS1 and ANO1, two units and at ONS1 AN01. has has resulted reassessing its conclusion in GL resulted in the staff reassessing GL 97-01 that cracking cracking of VHP nozzles nozzles is not an immediate immediate safety concern. Specifically. the findings indicate concem. Specifically, circumferential cracks indicate that circumferential cracks outside of the J-groove outside J-grOove welds can occur. occur, in contrast to an earlier earlier conclusion conclusion that the cracks cracks would be predominantly orientation. The fihdings predominantly axial in orientation. tiMings indicate indicate that cracking of the J-groove J-groove weld metal can precede precede cracking cracking of the base metal. metal. These findings raise questions questions regarding regarding the industry approach, developed in generic responses to GL 97-01, approach, developed 97-01. that utilizes PWSCC susceptibility modeling based on the base metal conditions and do not consider susceptibility consider those of the the weld metal. In addition, addition. the presence circumferential cracking presence of circumferential cracking at ONS3, where where only a small amount of boric acid residue indicated amount indicated a problem, adequacy of current info question the adequacy problem, calls into current visual examinations examinations for detecting detecting either circumferential cracking in VHP nozzles. This is either axial or circumferential especially significant if prior existing boric acid deposits on the RPV head mask the identification especially identification of new depoSits. Also, the presence of insulation on the RPV head new deposits. other impediments may head or other restrict an effective visual examination. As a remedial measure, the RPVhead remedial measure. RPV head may have to be be cleaned at aa prior outage for for effective identification of new deposits deposits from VHP nozzlenozzle cracking cracking ifif deposits cannot be discriminated from existing deposits new depOSits depoSits from otherother sources. However, the the NRC staff believes believes that boric acid deposits that cannot dispositioned as coming from another be dispositioned cannot be considered, as a conservative source should be considered. conservative assumption, to be from VHP nozzles, and appropriate corrective actions may be necessary.

appropriate necessary. In addition, the use of special tooling or procedures may be required to provide procedures assurance that the visual examinations will be effective in provide assurance in detecting the relevant conditions.

One function of VHP nozzles is to maintain the reactor coolant system pressure boundary. The The CRDM nozzles support and guide the control rods, and, therefore, are relied upon in shutting shutting down the reactor. Cracking of CRDM nozzles and welds isa is a degradation degradation of the reactor coolant coolant experience has shown that Alloy 600 is susceptible Industry experience system boundary. Industry susceptible to stress Further, the corrosion cracking. Further, the findings at ONS2 and ONS3 highlight the possible existence of S14D-04667 5140-04667 NRC007-2018 NRC007 *2018

BL 2001-01 BL2001-01 Page 55 of Page of1515 aa more more aggressive aggressive environment environment in the CRDM In the CRDM housinghousing annulus annulus following following through-wall through*walileakage; leakage; potentially highly concentrated borated primary water could become oxygenated in this annulus potentially highly concentrated borated primary water could become oxygenated in this annulus and possibly and possibly cause increa~ propensity cause increased propensity for forthe initiation of the initiation cracking and of cracking and higher highercrackcrack growth growth rates.

rates.

The cracking identified The cracking identified at at ONS2 ONS2 and and ONS3 reinforces the ONS3 reinforces importance of the importance conducting effective of conducting effective examinations of the RPV upper head area (e.g., visual examinations of the RPV upper head area (e.g., visual under*the--insulationexaminations under-the-insulation examinations of the ofthe penetrations pehetrationsfor evidence of for evidence of borated borated water leakage, or water leakage, examinations of volumetric examinations orvolumetric of the the CRDM CRDM nozzles), and nozzles), and using using appropriate appropriate NDE methods (such NDE methods (such as as PT, PT, UT, UT, and eddy-current testing) andeddy.current testing) to to adequately characterize adequately characterize cracks. cracks. Because Because of plant*specific design of plant-specific characteristics, there design characteristics, there is is no no uniform way uniform way to perform effective to perform examinations of visual examinations effective visual the RPV of the RPV head head at at PWR facilities. Some PWR facilities. Some plants plants have have the head insulation the head sufficiently offset insulation sufficiently offset fromfrom the RPV head the RPV head to permit an to permit an effective effective visual visual examination. Other plants have the insulation offset from the head but a contour matching examination. Other plants have the insulation offset from the head but in a contour matching that of that ofthe the head, head, requiring requiring special tooling and special tooling and procedures proc:eduresto to perform perform an effective visual an effective visual examination.

examination. Still other plants Still other plants have have insulation insulation directly directly adjacent adjacent to to oror attached attached to to the the RPV RPV head, head, potentially potentially requiring requiring the the removal removal of the insulation ofthe insulation to to permitpermit an an effective visual examination.

effective visual examination.

Several licensees have Several licensees recently performed have recently expanded VT-2 performed expanded examinations using VT*2 examinations using remote remote devicesdevices to inspect inspect between between the the RPV RPV head head and and thethe insulation.

insulation. One aspect of One aspect of conducting effective visual conducting effective visual examinations examinations that that isis common common to to all all PWR PWR plants plants is is the need to the need to successfully successfully distinguish distinguish boric boric acid acid deposits deposits originating originating with with VHP VHP nozzle nozzle cracking cracking from from deposits deposits that that are are attributable attributable to to other other sources.

sources.

For For boric boric acid deposits from acid deposits from CRDM CRDM nozzle nozzle cracks cracks to to be detectable at be detectable at the the outer outer surface surface of of the the .

RPV head, sufficient reactor cooIanthas to leak through the primary pressure boundary into RPV head, sufficient reactor coolant-has to leak through the primary pressure boundary into thethe annulus between the annulus between the CRDM CRDM nozzlenozzle and and the the RPVRPV head head base base metal, propagate up metal, propagate up the the annulus, annulus, and and finally emerge onto finally emerge onto the the outer surface of outer surface of the the RPV RPV head. head. SinceSince PWSCCPWSCC cracks cracks in in Alloy Alloy 600600 and and Alloy Alloy 182182 welds welds are are very tight. leakage very tight, leakage from axial cracks from axial cracks in in the nozzle and the nozzle and their their associated associated welds is welds is expected expected to to be be small.

small. In addition, possible In addition, possible restraint restraint of of pressure-induced pressure.inducedbendingof bending of circumferential cracks in CRDM nozzles could minimize the leakage available even from circumferential cracks in CRDM nozzles could minimize the leakage available even from CRDMCRDM nozzles with large nozzles with circumferential cracks, large circumferential cracKs, as evidenced by as evidenced by small boric acid small boric acid deposits identified at deposits identified at ONS3.

ONS3. As As described described in in Electric Power Research Electric Power Research Institute Institute .(EPRI)

(EPRI) Report Tp*1001491 , Part Report TP-1001491, Part2, 2, "PWR Materials Reliability "PWR Materials Program Interim ReliabilityProgram Alloy 600 Safety Interim Alloy.600 Safety Assessments Assessments for for US US PWR PWR PlantsPlants (MRP-44), Part 2: Reactor Vessel Top Head Penetrations (referred to as "the MRP-44, Part (MRP-44), Part 2: Reactor Vessel Top Head Penetrations" n (referred to as "the MRP-44, Part 2, 2, report"), the report), the majority majority of of CRDM CRDM nozzles nozzles are are installed installed into into the the RPVRPV head head with with an interference fit an interference fit at at room room temperature, temperature, with with 4343 plants plants having having specified specified interferenceinterference fit fit ranges ranges greater greater thanthan thosethose at at ONS ONS and ANOl. Should andAN01. Should these interference fits these interference persist at fits persist at plant operating conditions, plant operating conditions, they they could could provide an impediment to the flow of coolant leakage up the annulus and thereby limit the provide an impediment to the flow of coolant leakage up the annulus and thereby limit the amount amount of deposit available of deposit available on the RPV on the RPV head head for for detection detection by visual examination.

by visual examination.

The The recently identified CRDM recently identified CRDM nozzle degradation phenomena nozzle degradation phenomena raise raise several issues regarding several issues regarding the the approach taken in resolution approach in GL GL 97-01:

97*01:

S14D-04668 514D-04668 NRC007-2019 NRC007 *2019

BL 2001-01 Page60f15 Page 6 of 15 (1) Cracking of Alloy 182 weld metal has been identified CRDM nozzle J-groove welds for identified in CROM for the first time. This finding raises an issue regarding the adequacy adequacy of cracking cracking susceptibility models susceptibility models based based only on the base metal conditions.

conditions.

(2) The identification of cracking at AN01 AN01 raises an issue regarding the adequacy of the the susceptibility model. ANO1 industry's GL 97-01 susceptibility AN01 cracking was w~s predicted to be more than 15 effective full power poweryears years (EFPY)

(EFPy) beyond beyond January 1, 1997 1997,,frOm from reaching the same same conditions as the limiting plant, based based on the susceptibility susceptibility models models used by the industry industry to address base metal cracking in response to Gl97~1. GL 97-01.

(3) Circumferential cracking of CRDM Circumferential CROM nozzles, located outside of any structural retaining retaining welds, has been identified for the first time. This finding raises concerns concerns about the the potential for rapidly propagating potential propagating failure of CRDM nozzles and control rod ejection, CROM nozziesand ejection, causing a loss of coolant accident accident (LOCA).

(LOCA).

(4) Circumferential crackingftom Circumferential cracking from the CRDM nozzle OD 00 to the.ID has been identified forthe totheJD for the first time. This finding raises concerns about increased consequences consequences of secondary secondary effects effects of leakage from relatively benign axial cracks.

(5) Circumferential cracking of CROMnozzies Circumferential CRDM nozzles was identified by the presencepresence of relatively relatively small amounts of boric acid deposits. This finding increases increases the need for more effective effective inspection methods to detect detect the presence bf degradation presence "of degradation inCRDM in CRDM nozzles before the the nozzle integrity integrity is compromised.

compromised.

After the initial finding of significant circumferential circumferential cracking cracking at ONS3, the NRC held a public public meeting with the EPRI Materials Reliability Program Program (MRP) on April 12, 2001, 2001, to discuss CRDM nozzle nozzle circumferential cracking cracking issues. During the meeting, the industry representatives representatives indicated that they were developing a genericgeneric safety assessment, recommendations recommendations for revisions of near-term inspections, inspections, and long-term inspection and flaw evaluation evaluation guidelines.

guidelines.. On 18, 2001, the MRP submitted the MRP-44.

May 18.2001. MRP-44, Part 2. 2, report to provide an interim safety assessment for PWSCC of Alloy 600 VHP nozzles nozzles and Alloy 182J-groove 182 J-groove welds in PWR plants.

On June 7, 2001, 2001, the NRC held a public meeting meeting at which the MRP MRP provided provided initial responses to questions on questions on the the MRP-44, MRP-44, Part 2, report that the NRC staff had identified and transmitted to the the MRP on MRP May 25, on May 2001.

25, 2001.

The approach taken in the MRP-44, Part Part 2, report uses useS an assessment assessment of the relative relative susceptibility of each PWRPWR to OD-initiated aD-initiated or weld PWSCC based on the operating time and temperature of the penetrations. Based upon this simplified model. model, provided provided in Appendix B of the MRP-44, Part 2, report, each PWR plant was ranked report,each MRP according to the operating ranked by the MRPaccording operating time in EFPY required for the plant to reach an effective effective time-at-temperature time-at-temperature equivalent to ONS3 at the the time the above-weld circumferential cracks were identified above-weld circumferential identified in early 2001.

2001. To address the the experience at ONS, the report recommended experience atONS, recommended that plants ranked within 10 EFPY of ONS3 and having fall 2001 2001 outages should perform a visual inspection inspection of the RPV top head capable of detecting detecting small amounts that observed amounts of leakage similar to that observed at the Oconee Oconee units and AN01.ANOI.

BL 2001-01 Bl2001-01 SI4D-04669 S140-04669 NRC007 -2020 NRC007-2020

Page Page 770fof 15 The NRC staff questions to the MRP provided questions staffprovlded aspects of the MRP-44, MRP on various aspects MRP-44, PartPart 2, report in June 22, 2001; the letter dated June22,2001; in aa letter the MRP provided responses in a letter provided responses dated June 29, 2001.

letter dated 2001.

These These questions addressed aspects of questions addressed of the proposed proposed 'industry treatment that industry treatment that the the NRC staff did staff did not agree with. Two not agree Two specific areas of specific areas concern are (1) the ofconcem finding that nozzle the finding nozzle leaks are detectable on detectable on all vessel heads, .and (2) the lack consideration of an applicable lack of consideration applicable crack growth rate for the VHP VHP nozzle nozzle cracking a-acking situation (including (including aa conclusion in in the MRP MRP responses that the the appropriate appropriate crack crack growth rate for OD growth rate cracking of VHP 00 a-acking nozzles is represented VHP nozzles represented by data from aa by data primary water primary water environment). nOzzle leaks issue of detectibility of nozzle environment). The issueofdetectibility leaks in particular plant is in any particular difficult to address plant-specific as-built address due to a need for plant-specific such as geometries, such as-built geometries, as measured dimensions on CRDM penetrations to characterize nozzles and RPV penetrations CROM nozzles interference fit characterize the interference population population for a particular head. In particular RPV head. In addition, there there is aa need provide aa sufficiently detailed need to provide model of the RPV head and expected ofthe expected through-wall through-wall crack crack characteristics, characteristics, suchsuch. as surface surface roughness roughness and crack crack tightness, to provide assurance assurance that any nozzles nozzles with through-wall cracking will provide sufficient cracking leakage to the RPV .head sufficient leakage head surface surface such deposits of such that residual deposits boric boric acid will provide detectable condition provide a detectable COndition for the visual examination. inability to provide examination. An inability provide assurance assurance of a detectable detectable residual residual deposit existing boric acid deposits discriminate prior existing deposit or to discriminate deposits caused non-safety-slgnlficant sources from boric acid caused by non-safety-significant acid deposits caused by CROM CRDM nozzle nozzle effectiveness of visual examinations.

cracking could limit the effectiveness cracking Because visual examination examination of the RPV head or volumetric volumetric examination examination of the VHP nozzlesnozzles periodically (generally occurs only periodically at a scheduled refueling outage), the issue of crack growth (generally atascheduled in VHP nozzles rate in nozzles is an important consideration consideration in providing assurance assurance that VHP nozzles will thatVHP maintain their structural integrity between examinalion theirstrudural examina~tion opportunities.

opportunities. In In particular, crack growth should be low enough to ensure that Vf-:lP VHP nozzles which are determined determined to be unflawed during during an examination examination do not have critical flaw sizes prior to the next scheduled scheduled examination.

From the results of the susceptibility ranking model model proposed proposed in in Appendix B B to MRP-44, MRP-44, Part 2, the population of PWR plants can be divided into several subpopulationssubpopulations with similar characteristics:

those plants plants which have demonstrated demonstrated the existence existence of PWSCC PWSCC in in their VHP nozzles nozzles (through the detection of boric acid deposits) and for which cracking cracking can be expected expected to recur and affect affect additional VHPs; considered as having a high susceptibility to PWSCC based those plants which can be considered upon a susceptibility ranking of less than 5 EFPYfromEFPY from the ONS3 condition; those plants which can be considered as having a moderate susceptibility susceptibility to PWSCC PWSCC

based based upon a susceptibility ranking of more than 5 EFPY but less than 30 EFPY from the the ONS3 condition; and the balance of plants which can be considered as having low susceptibility based upon a susceptibility ranking of more than 30EFPY 30 EFPY from the ONS3 condition.

Although the industry susceptibility ranking model has has limitations, such as large uncertainties uncertainties and nono predictive capability, the model does provide aa starting point for assessing the potential nozzle cracking in for VHP nozzle in PWR plants.

BL 2001-01 BL 5140-04670 SI4D-04670 NRC007-2021 NRC007 -2021

Page Page 88 of 15 The following paragraphs characterize following paragraphs dlaracterizethe gradation of inspection effort for the subpopulations the gradation subpopulationsof of plants noted above. Nevertheless, Nevertheless. addressees addressees should should bebe cognizant cognizant of extenuating circumstances at their respective circumstances respective plant(s) that would suggest suggest a need need for more more aggressive aggressive practices to provide inspection practices inspection provide an appropriate appropriate level of confidence confidence in VHP nozzle nozzle integrity.

Integrity. In In addition, since inspection inspection and repair activities can potentially potentially result in large personnel exposures, large personnel licensees should ensure licensees activities related to the inspection ensure that all activities inspection of VHP VHP nozzles nozzles and the repairrepair of identified identified degradation degradation areare planned planned and and implemented implemented to.keep to keep personnel exposures as low as personnel exposures as reasonably achievable achievable (ALARA),

(ALARA), consistent With the NRC consistent with ALARA policy.

NRCALARA For thethe subpopulation subpopulation of plants considered considered to have have aa low susceptibility susceptibility to PWSCC, based based upon upon a susceptibility susceptibility ranking ranking of more than 30 EFPY from the ONS3 ONS3 condition, the anticipated anticipated low low likelihood of PWSCC degradation likelihoOd degradation at these facilities indicates indicates that enhanced examination beyond enhanced examination beyond the current requirem~nts requirements is not necessary necessary at the present time because presenttime because there is aa low likelihood likelihood that the enhanced examination would provide additional enhanced examination additional evidence of the propensity propensity for PWSCC in VHP nozzles. .

For the subpopulation subpopulation of plants considered have aa moderate considered to have moderate susceptibility to PWSCC PWSCC based upon upon a susceptibility susceptibility ranking of more than 5 EFPYbut EFPY.but less than 30 EFPY EFPY from the ONS3 ONS3 condition, condition, an effective visual examination, examination, at a minimum, of 100% 100% of the VHP nozzles ofthe nozzles that is is capable capable of detecting detecting and discriminating small amounts amounts of boric acid add deposits from VHP nozzle nozzle leaks, such as were identified identified at ONS2 and QNS3, ONS3, maybe sufficient sufficient to provide reasonable reasonable confidence that PWSCC degradation confidence degradation would be identified identified prior to posing poSing an undue risk. This This effective effective visual examination examination should not not be compromised comprolnised by the presence.of presence of insulation, existing existing deposits on the depoSits RPV head, or other theRPV other factors factors that could interfere ~ithwith the detection ofleakage.

of leakage.

For the subpopulation subpopulation of plants considered considered to have a high susceptibility to PWSCC PWSCC based upon a susceptibility ranking of less than 5 EFPY from the ONS3 condition, the possibility of VHP VHP nozzle cracking at one of these facilities indicates the need to use a qualified qualified visual examination examination of o.f 100% of of the VHP nozzles. This qualified qualified visual examination examination should be able to reliably detect and accurately characterize leakage accurately characterize leakage from cracking cracking inin VHP nozzles considering considering two characteristics. One characteristics. One characteristic plant-specific demonstration characteristic is a plant-specific demonstration that any VHP nozzle nozzle exhibiting through-wall leakage to the RPV head surface (based through-wall cracking will provide sufficient leakage on the as-built configuration of the VHPs). Secondly, similar to the effective effective visual examination examination for moderate susceptibility susceptibility plants, the effectiveness of the qualified visual examination should not be compromised compromised by the presencepresence of insulation, existing deposits on the RPV head, or other factors that could interfere with with the detection of leakage. Absent the use of a qualified qualified visual qualified volumetric examination examination, a qualified examination of 100% of the VHP nozzles nozzles (with a demonstrated capability to reliably detect cracking on the 00 OD of a VHP nozzle) may be be appropriate to provide appropriate provide evidence evidence of the structural structural integrity of the VHP nozzles.

subpopulation of plants which have already identified For the subpopulationof identified the existence of PWSCC in the in the CRDM nozzles (for example, through the detection of boric acid deposits), deposits), there is a a sufficient sufficient likelihood that the cracking of VHP nozzles likelihOOd nozzles will continue continue to occur as the facilities continue to qualified volumetric examination of 100% of the VHP nozzles (with aa operate. Therefore, a qualified demonstrated capability to reliably detect cracking on the 00 OD of the VHP nozzle) may be be appropriate to provide evidence of the structural integrity of the VHP nozzles.

5140-04671 S14D-04671 NRC007-2022 NRC007 -2022

BL2001-Q1 BL 2001-01 Page90f Page 9 of 15 developed a Web page to keep the public informed of generic The NRC has developed generic activities on PWR Alloy 600 weld cracking (http:/lwww.nrc.gov/NRC/REACTORIALLOY-600/index.html).

(http:/twww.nrc.govINRC/REACTORlALLOY-600/index.html).This This page provides links to information information regarding regarding the cracking identified to date, along with documentation of NRC interactions documentation interactions with industry (industry submittals, meeting notices, presentation materials, and meeting summaries). The NRC NRC will continue to update update this this

. Web page page as new information becomes available.

Apolicable Regulatory ReMuirements Applicable ReQuirements Several provisions of the NRC regulations licenses (Technical operating licenses regulations and plant operating Specifications)

Specifications) pertain pertain to the issue of VHP nozzle nozzle cracking. The generalgeneral design design criteria (GOG)

(GDC) nuclear power plants (Appendix A to 10 CFR Part 50), or, as appropriate, similar for nuclear requirements in the the licensing licensing basis fora for a reactor facility, the requirements requirements of 10 CFR 50.55a, and the quality assurance assurance criteria of Appendix Appendix B to 10 CFR Part 50 provide provide the bases and requirements for NRC staff assessment of the potential for and consequences consequences of VHP nozzle nozzle cracking.

The applicableGOC applicable GDC include GDC GOC 14, GOC GDC 31,31, and GOC GDC 32. GOCGDC 14 specifies that the reactor coolant pressure (RCPB) have an ~xtremely pressure boundary (RCPB).have extremely low probability probability Ofof abnormal abnormal leakage, of rapidly propagating failure, and of gross rupture; the the presence presence of cracked and and leaking VHPVHP

nozzles is not consistent with this GOC.

nozzles GDC. GOC GDC 31 speCifies specifies that the probability probability of rapidly rapidly propagating propagating fracture of the RCPB be minimized; minimized; methe presence of cracked cracked and leaking VHP leaking VHP nozzles is not consistent with this GOC. GDC. GOC GDC 32*specifies 32 specifies that components components which are part of the RCPB have the capability of being periodicallyperiodically inspected to assess their structural and and leaktight leaktight integrity; integrity; inspection inspection practices that do not permit reliable detection practices thaldo detection of VHP nozzle nozzle cracking cracking are not consistent consistent with this GOC. GDC.

NRC regulations at 10 CFR 50.55a 50.553 state that ASME Class 1I components (which include VHP VHP nozzles) must meet the requirements requirements of Section XI Xl of the ASME Boiler and Pressure Vessel Code. Table IWA-2500-1 IWA-2500-1 of Section XI of the ASME Code provides requirements provides examination requirements for VHP nozzles nozzles and references references IWB-3522 IWB-3522 for acceptance standards. IWB-3522.1(c)

IWB-3522.1 (c) and (d) specify that conditions requiring correction include include the detection detection of leakage from insulated components components and discoloration or accumulatedaccumulated residues on the surfaces surfaces of components, insulation, or floor areas which may insulation, .may reveal evidence of borated borated water leakage, with leakage leakage "the through-wall leakage defined as "the leakage that penetrates the pressure retaining membrane."

membrane,."

Therefore, 10 CFR 50.55a, through its its reference reference to the ASME Code, does not permit through-wall cracking cracking of VHP nozzles.

noZZles.

For through-wall through-wall leakage identified by by visual examinations examinations in accordance accordance with the ASME Code, acceptance standards acceptance standards for the identified identified degradation degradation are provided provided in IWB-3142.

IWB-3142. Specifically, Specifically, supplemental examination (by surface or volumetric examination), examination), corrective corrective measures or repairs, analytical analytical evaluation, evaluation, and replacement replacement provide provide methods for for.determining determining the acceptability acceptability of degraded degraded components. .

IX of Appendix Criterion IX Appendix B to to 10 CFR CFR Part 50 states that special processes, includingincluding nondestructive testing, shall be controlled and accomplished nondestructive accomplished by qualified qualified personnel personnel using using S14D-04672 8140-04672 NRC007-2023 NRC007 -2023

BL2001-01 BL 2001-01 Page 10 of 15 qualified procedures qualified procedures in accordance accordance with applicable applicable codes, standards, specifications, specifications, criteria, and other special special requirements.

requirements. Within the context of providing providing assurance assurance of the structural integrity of VHP nozzles, special special requirements for visual examination examination would generally require the the examination method.

use of a qualified visual examination method. Such a method method is one that a plant-specific plant-specific analysis has demonstrated demonstrated will result in sufficient sufficient leakage to the RPV head head surface surface for aa through-wall crack In a VHP nozzle,and through-wall nozzle, and that the resultant leakage provides a detectable detectable deposit analysiS would .have on the RPV head. The analysis have to consider, for example, the as-built configuration of the VHPs and the capability to reliably characterize the source of the reliably detect and accurately characterize the leakage, considering considering the presence of insulation, preexisting preexisting deposits deposits on theRPV the RPV head, and other factors that could interfere with the detection detection of leakage. Similariy, Similarly, special requirements for volumetric examination examination would generally require the use of aa qualified qualified volumetric examination method, for example, one that has has a demonstrated demonstrated capability to reliably detect detect cracking on the the 00 OD of the VHP noZZle nozzle above the J-groove weld.

Criterion V of Appendix Appendix B to 10 CFR Part 50 states that activities affecting quality shall be be prescribed by documented prescribed documented instructions, procedures, or drawings, draWings, of a type appropriate to the the circumstances and shall be accomplished circumstances accomplished in accordance accordance with these instructions, Instructions, procedures, or drawings. Criterion Criterion V further states that instructions, procedures, .or thatinstructions, draWings shall.include or drawings shall include appropriate quantitative or qualitative appropriate quantitative qualitative acceptance acceptance criteria for determining determining that important activities important activities have been satisfactorily satisfactorily accomplished.

accomplished. Visual and volumetric volumetric examinations of VHP nozzlesnozzles are activities that should be documented documented in accordance accordance with these requirements.

requirements.

"J;;,

Criterion Criterion XVI of Appendix B to 10 CFR Part 50 states that measures measures shall be established established to assure that conditions adverse adverse to quality are promptly identified and corrected. For Significant promptly identified significant conditions adverse to quality, the measures taken shall include root cause determination determination and corrective action to preclude repetition corrective repetition of the adverse conditions. For cracking cracking of VHP nozzles, the root cause determination determination is important to understanding understanding the nature of the degradation present degradation present and the required actions to mitigate *future future cracking. These actions could include proactive proactive inspections and repair of degraded degraded VHP nozzles.

Plant technical specifications pertain technical specifications pertain to the issue of VHP nozzle nozzle cracking require cracking insofar as they require no through-wall reactor through-wall reactor coolant system leakage.

leakage.

Requested Information Requested Information This bulletin bulletin requests addressees addressees to submit information.

information. Addressees Addressees who choose to utilizeutilize the the analyses provided provided in the MRP-44, MRP-44, Part 2, report or similar analyses analyses need to consider consider the NRCNRC staff questions questions relative to this report (provided to the MRP by letter dated 22, 2001) when dated June 22,2001) preparing preparing their plant-specific plant-specific responses to the requested information.

information. Addressees Addressees should note note that the NRC staff has found that the industry response response to these questions (provided by letter letter

. dated June 29, 2001) does not provide aa sufficient basis for resolving June 29,2001) resolving the relevant technical issues is.sues and that additional information will be necessary additional necessary to support support the plant-specific plant-specific evaluations.

Addressees are requested requested to provide the requested requested information within 30 days of the date of this this bulletin (except for Item Item 5).

BL 2001-01 BL 2001-01 S14D-04673 5140-04673 0 NRC007-2024 NRC007 -2024

Page Page 11 of 15 of15

1. addressees are requested All addressees requested to provide provide the the following following information:

infomlation:

a. the plant-specific the susceptibility ranking for your plant-specific susceptibility plant(s) (including your plant(s) (including all all data used used to determine each determine each ranking) using using the the PWSCC PWSCCsusceptibility described in susceptibility model described Appendix B Appendix B to to the MRP-44, MRP-44, Part 2, report;

b. description of the a description the VHP nozzles nozzles inin your your plant(s), including the number, type, inside plant(s), including Inside and outside diameter, materials materials ofof construction, construction, and minimum distance between and the minimum between VHP VHP nozzles; nozzles;

c. aa description description of oftha the RPV headhead insulation insulation type and and configuration; configuration;

d. aa description description ofthe nozzle and RPV head inspections (type, scope, qualification of the VHP nozzle qualification requirements, requirements, and acceptance acceptance criteria) that have been perfomled plant(s) in the performed at your plant(s) the past 4 years, and and the findings. Include Include aa description limitations (insulation or deSCription of any limitations or accessibility of the bare metal of the RPV head other impediments) to accessibility other head for visual examinations;

e. a description of the configuration configuration of the missile housings and their CRDM housings missile shield, the CRDM their support/restraint components, structures, and cabling from the top of support/restraint system, and all components, of the RPV head up to the missile shield. Include the elevations elevations of these these items items relative to the bottom of the missile shield.

2. If your plant has previously previously experienced experienced either leakage from or cracking cracking in VHP nozzles, addressees are requested to provide the following addressees following Information:

a. a description of the extent of VHP nozzle nozzle leakage and cracking cracking detected detected at youryour plant, including location, size, and nature of each crack detected; including the number, location, detected;

b. descrption of the additional a description additional or supplemental inspections (type, scope, qualification supplemental inspections qualification requirements, requirements, and acceptance acceptance criteria), repairs, and other corrective actions you have taken in response identified cracking to satisfy applicable response to identified regulatory requirements; applicable regulatory

c. inspections (type, scope, qualification your plans for future inspections qualification requirements, and acceptance criteria) and the schedule; acceptance

d. basis for concluding that the inspections identified in 2.c will assure that regulatory your basis.for requirements are met (see Applicable requirements Regulatory Requirements Applicable Regulatory Requirements section). Include Include thethe following specific information in this discussion:

If your future inspection plans do not include performing (1) If performing inspections before before December 31, 2001, provide your basis for 31, 2001, for concluding that the regulatory regulatory Applicable Regulatory requirements discussed in the Applicable Regulatory.Requirements Requirements section will continue to be met untiluntil the inspections are performed.

If your future inspection plans do not (2) If include volumetric examination of all VHP not indude VHP

.provide your basis for concluding that the regulatory reqUirements nozzles, provide requirements discussed in the Applicable Regulatory Requirements section will be satisfied.

Requirements section S14D-046 7 4 5140-04674 NRC007-2025 NRC007 -2025

BL2001-01 BL 2001-01 Page 12 of 15 15

3. Ifthe susceptibility ranking for your plant is within 5 EFPY of ONS3, addressees are If requested provide the following information:

requested to provide information:

a. your plans for future inspections (type, scope, qualification inspections (type, qualification requirements, and acceptance acceptance criteria) criteria} and the schedule;

b. your basis for concluding concluding that the inspections identified identified .in in 3.a. will assure that requirements are met (see Applicable Regulatory regulatory requirements Regulatory Requirements Requirements section).

Include the following specific information information in in this discussion:

(1) IfIfyour Mure (1) future inspection inspection plans do not include performing Inspections Inspections before before .

December December 31, 31, 2001, 2001, provide your basis for concluding concluding that the regulatory Applicable Regulatory Requirements requirements discussed in the Applicable requirements Requirements section will will performed.

inspections are performed.

continue to be met until the inspections (2) If inspection plans include only visual inspections, discuss the Ifyour future inspection the

.corrective corrective actions actions that will be taken, taken, including alternative alternative inspection inspection methods (for(for example, example, volumetric volumetric examination),

examination}, if ifleakage leakage is detected.

4. Ifthe susceptibility If susceptibility ranking for your plant is greater than 5 EFPY and less than 30 EFPY of ONS3, addressees addressees are requested requested to provide the following information:

~.

a. your plans for future future.inspections inspections (type, scope, qualification qualification requirements, requirements, andand acceptance criteria) and the schedule; acceptance

b. your basis for concluding concluding that the inspections identified in in 4.a will assure that regulatory requirements are met (see Applicable Regulatory Requirements ApplicableR9gulatory Requirements section). Include Include the the following specific information in in this discussion:

(1) IfIfyour future inspection (1) inspection plans do not include a qualifiedqualified visual examination examination at the the next scheduled scheduled refueling outage, provide your basis for concluding that the .,

requirements discussed in regulatory requirements in the Applicable Requirements Applicable Regulatory Requirements section will continue to be met until the inspections are performed.

(2) The corrective actions that will be taken, including including alternative inspection methods inspection methods (for example, volumetric .examination),

examination), ififleakage leakage is detected.

detected.

5. Addressees Addressees are requested to provide the following information information within 30 days after plantplant restart following the next refueling outage:

a.

a. aa deSCription description of of the the extent extent of VHP nozzle of VHP nozzle leakage leakage andand cracking detected at your plant, cracking detected plant, including the number, location, size, and nature of each including each crack detected; S14 ID.04675 5140-04675 NRC007-2026 NRC007 -2026

BL2001-01 BL 2001-01 Page 13 Page 130f of 15

b. cracking is identified, if cracking identified, a description description of thethe inspections inspections (type, scope, qualification qualification requirements, and acceptance requirements, acceptance criteria),

criteria), repairs, and other corrective actions other corrective have actions you have taken to satisfy taken satisfy applicable applicable regulatory regulatory reqUirements.

requirements. This information information Is requested only is requested are any changes if there are changes from priorprior information information submitted accordance with this submitted in accordance this bulletin.

bulletin.

Required Response ReQuired Response In accordance accordance with 10 10 CFR CFR 50.54(f),

5O.54(f), in order order to determine determine whether any any license license should should be be suspended, or revoked, each modified, suspended, each addressee required to respond addressee is required described below.

respond as described This information information is is sought to verify compliance with the current licensee compliance verify licensee licensing basis current licensing basis for the forthe facilities covered by this bulletin.

facilities covered bulletin.

Within 30 days days of the date of this bulletin, addressee is required to submit a written bulletin, each addressee indicating (1) whether the requested response indicating response information will be submitted and (2) whether the requested information the requested information requested submitted within the requested Information will be submitted requested time period. Addressees who period. Addressees who choose not to submit the requested requested information, or are are unable to satisfy satiSfy the requested requested completion date, must describe in their response completion response any alternative courSe of action alternative course action they propose propose to take, acceptability of the proposed alternative including the basis for the acceptability take,including alternative course of action.

required written The required written response should be addressed Nuclear Regulatory addressed to the U.S. Nuclear Commission, ATTN: Document Commission, Document Control Desk, Wakhington.

Wathington, DC 20555-0001.

20555-0001, under oath or affirmation under Section 182a of the Atomic Energy Act of 1954, as amended, provisions of Section under the provisions and 10 10 CFR 50. 54(f). In addition, addition, submit a copy of the response to the appropriate appropriate regional administrator.

Information Reeuest Reasons for Information Request Through-wall cracking of VHP nozzles Through-wall nozzles violates NRC regulations and plant technical VHP nozzles can pose a safety risk if permitted to Circumferential cracking of VHPnozzles specifications. Circumferential progress to the pOintpoint that nozzle nozzle integrity is in question and the risk of a loss of coolant coolant accident accident or probability ofa nozzle ejection increases. This information request is necessary of a VHP nozzle necessary to assessment of plant-specific permit the assessment compliance with NRC plant-specific compliance NRC regulations. This information will also be used by the NRC staff to determine determine the need for and to guide the development development of additional regulatory actions to address cracking in VHP nozzles. Such regulatory actions could requirements for augmented inspection programs under 10 CFR 55a(g)(6)(ii) include regulatory requirements 55a(g)(6)(ii) additional generic communication, or additional communication. .

Related Generic Communications Communications

• Information Circumferential Cracking Information Notice 2001-05, "Through-Wall Circumferential Cracking of Reactor Pressure Vessel Head Head Control Mechanism Penetration Nozzles at Oconee Nuclear Control Rod Drive Mechanism Nuclear Station, UnitUnit 3," April 30.

30, 2001. ML011160588]

2001. [ADAMS Accession No. ML011160588]

S14D-04676 5140-04676 NRC007-2027 NRC007 -2027

BL2001-01 BL.2001-01 Page 14 Page 14 of of 15

• • Generic 97-01, "Degradation Generic Letter 97-01, "Degradation of Control Rod Rod .Drive Mechanism Mechanism Nozzle Nozzle and Other Other Vessel Closure Head Penetrations," April 1, 1997.

Closure Head 1997.

" Information Notice Information 96-11, "Ingress of Demineralizer Notice 96-11. Demineralizer Resins Resins Increases Increases Potential Potential for Stress Stress Cracking of Corrosion Cracking Corrosion Mechanism Penetrations,"

Control Rod Drive Mechanism of Control February 14, 1996.

Penetrations," February 1996.

Information Notice Information 90-10, "Primary Water Notice 90-10, Stress Corrosion Water Stress Corrosion Cracking Cracking of INCONEL 600,"

of(NCONEL 600,"

23, 1990.

February 23,1990.

February Generic Corrosion of Letter 88-05, "Boric Acid Corrosion Generic Letter of Carbon Reactor Pressure Carbon Steel Reactor Pressure Boundary Components in Components in PWR Plants," March March 17,1988.

17, 1988.

NUREG/CR-6245, "Assessment of Pressurized NUREG/CR-6245, Pressurized Water Reactor ControlControl Rod Drive Drive Mechanism Nozzle Cracking," October Mechanism October 1994.

Backfit Backfit Discussion Under provisions of Section 182a of the Atomic Under the provisions 1954, as amended, and AtomiC Energy Act of 1954,.asamended,and 10 CFR 50.54(f),

5O.54(f), this generic generic letter transmits an information request for the purpose verifying purpose of verifying compliance compliance with existing applicable requirements (see the Applicable applicable regulatory requirements Applicable Regulatory Requirements section Requirements section of this bulletin). Specifically, the requested requested information information will enable thethe NRC staff to determine NRC determine whether whether current inspectiort inspectiort practices for the detection detection of cracking in inthe the VHP nozzles VHP nozzles at reactor facilities provide provide reasonable reasonable confidence that reactor reactor coolant coolant pressure boundary boundary integrity is being maintained.

maintained. The requested information will also enable the NRC staff staff to determine determine whether addressee addressee inspection practices need to be augmented to ensure that the* the safety significance of VHP nozzle cracking remains low. No backfit VHPnozzie backfit is either intended or approved approved by the issuance issuance of this bulletin, and the staff has not performed a backfit analysis.

Federal Register RegisterNotification Notification A notice of opportunity opportunity for public comment on this bulletin bulletin was not published in in the the FederalRegister Federal Registerbecause the NRC staff is requesting informationinformation from power licensees power reactor licensees on an expedited basis for the purpose of of asses assessing applicable sing compliance with existing applicable regulatory requirements and the need for subsequent regulatory action. This bulletin was prompted by the discovery of circumferential circumferential cracking in in CRDM nozzles (above the nozzle-to-nozzle-to-vessel head weld) from the 00 OD to the 10 IDand cracking in in the J-groove J-groove weld metal itself. Both of phenomena have not these phenomena not been previously identified identified in in PWRs. As the resolution of this matter progresses, the opportunity for public involvement involvement will be provided; provided.

S14D-04677 5140-04677 NRC007-2028 NRC007 -2028

BL2001-01 BL 2001-01 Page Page 15 15 of 15 of15 Paperwork paperwork Reduction Reduction Act Statement Statement This bulletin contains information infonnation collections collections that areare subject subject to to the Paperwork Paperwork Reduction Act of 1,995 (44 U.S.C.

1995 3501 et U.S.C.3501 et seq.) These information collections Theseinfonnation collections were were approved approved by the Office Office of Management Management and and Budget, approval approval number number 3150-0011.

3150-0011.

The burden burden to the public for these mandatory mandatory information infonnation collections collections isis 140 hours0.00162 days <br />0.0389 hours <br />2.314815e-4 weeks <br />5.327e-5 months <br /> hours per per response, including including the time time for reviewing reviewing instructions, searching searching existing data sources, gathering gathering and maintaining maintaining thethe data data needed, needed, and completing completing and and reviewing revieWing the information infonnatlon collection. Send collection. Send comments comments regarding regarding this burden estimate estimate or or on on any other other aspect of these information theseinfonnation collections, including suggestions for reducing the burden, to the Records Records Management Management Branch Branch (T-6 E6), U.S. Nuclear Nuclear Regulatory Regulatory Commission, Commission, Washington, DC 20555-0001, 20555-0001, or by Internet Internet electronic electronic mail to BJS1@NRC.GOV; BJS1 @NRC.GOV; and and to the Desk Officer, Office of Information Infonnation and Regulatory Affairs, NEOB-10202 Regulatory NEOB-10202 (3150-0011), Office Office of Management Management and Budget, Washington, Washington, DC 20503.

Public Protection Protection Notification Notification If If a means used to impose impose an information infonnation collection does not displaydisplay aa currently currently valid OMB OMB control number,the number, the NRC NRC may not conduct or sponsor, and a person is not required to respond to, the information colledion.

theinfonnation collection.

If you have any questions If questions about this matter, please contact the technical contact listed below or appropriate Office of Nuclear the appropriate Nuclear Reactor Reactor Regulation Regulation (NRR) project manager.

/RAJ Matthews, Director David B. Matthews, Director Division of Regulatory Improvement DiviSion Improvement Programs Programs Office Office of Nuclear Reactor Regulation Nuclear Reactor Regulation Technical

Contact:

Allen L Hiser, Jr A1lenL Jr.,.* NRR 301-415-1034 301-415-1034 alhl@nrc.gov E-mail: alh1@nrc.gov Manager.

Lead Project Manager: Jacob I.I. Zimmerman, Zimmennan, NRR NRR 301-415-2426 301-415-2426 E-mail: jiz@nrc.gov

Attachment:

Schematic Figure of Typical CRDM Nozzle Nozzle Penetration S14D-04678 S140-04678 NRC007-2029 NRC007 -2029

Attachment Attachment Bl2001-01 BL 2001-01 Page 1 of 1 SA-182 F304 ERNiCr-3 SB-167 UNS.N06600 (Alloy 82)

(Alloy 600) N Outer Surface of RPV Head 4.-*" RPV Head RPVHead (SA-533 Gr. B Cl.CI. 1)

Shrink Fit -~-'I Surface cl RPV Head (Stainless Steel Cladding)

J-Groove Weld

.EniCrFe-3 (Alloy 182)

Schematic Figure of Typical Schematic Typical CRDM CRDM Nozzle Nozzle Penetration Penetration 514D-04679 S14D-04679 NRC007-2030 NRC007 -2030

/Z&AS Davi/Sew Davis-Besse Nudjsar PwrSSlalion FistEnergy .

Nur::Jeai Power 5501.

5501 NOIfh Oak Harbor. OWo Route 2 Slats Route NorUt Stat Ohio 43449-870 43449-6760

- - Guy G Campbell n - 419-321-8588

'."*iL V'ice Pa*4nt -.Nuedear Fax: 419-321-8337 U.S. NRC In NRC In re DAVID GEISEN.?

(+ (\ ExhIbIt GEISENV'LJ&..

i 'f\

0- t t Exhibit #

VI

1. ---I.J__

September September 4, 4, 2001 1A-05-052 Docket # 1A-OS-OS2 -

Docket Number 50-346 50-346 Date Marked for Date Markedfof ijoI!d1.-, 2008 (Tr.p 2008 (Tr. p.. 22'7 )

)

License Number NPF-3 Date Offered in J1jL..

inEv: _4 , 2008 (Tr. p.X (Tr.I) l (P )

Through witnesstganel:--Jf0~I+-1 Thrugh Witness!Ranel: ,)/ ;".-Pr &_ _ _ _ _-

Serial Number 2731 Action:

Actf ~ REJECTED AI§MDT~b REJECTED WITHDRAWNWITHDRAWN E4-L.

Date: 124i. 2D Date- 2008(Tr. p.i) LA.!L' )

(fr. p>i U.S. Nuclear Nuclear Regulatory Commission Commission Attention: Document Control Desk Desk Washington, Washington, D.C. 20555-0001

Subject:

Response to NRC Bulletin 2001-01, "Circumferential Cracking of Reactor

Subject:

Response to NRC Bulletin 2001-01, "Circumferential Cracking of Reactor Pressure Vessel Head Penetration Penetration Nozzles" ,

Ladies and Gentlemen:

3, 2001, the Nuclear On August 3,2001, Nuclear Regulatory Regulatory Commission Commission (NRC) issued NRC aulletin Bulletin 2001-01, "Circumferential Cracking of Reactor Pressure Vessel Head Penetration 2001-01, "Circumferential Penetration Nozzles."

Nozzles." The Bulletin Bulletin requested information structural integrity of the information regarding the structural reactor pressure vessel head penetration penetration (V-P)

(VHP) nozzles, including the extent of nozzle leakage and cracking that has been found to date, inspections and repairs that have been leakage arid completed completed to satisfy applicable applicable regulatory requirements, requirements, and the basis for concluding that plans for future inspections inspections will ensure compliance compliance with applicable applicable regulatory regulatory requirements.

requirements.

The Davis-Besse Nuclear Nuclear Power Station (DBNPS) has scheduledscheduled VHP inspections during the upcoming spring 2002 refueling outage. The FirstEnergy Nuclear Operating Nuclear Operating Company (FENOC) provides information for the DBNPS in response to provides the attached infonnation to 2001-01. .

NRC Bulletin 2001-01.

DOCKETED DOCKETED USNRC September 9, September 9, 2009 (11:00am) 2009 (11 :OOam)

OFFICE OF SECRETARY OFFICE OF SECRETARY RULEMAKINGS RULEMAKINGS AND AND ADJUDICATIONS ADJUDICATIONS STAFFSTAFF NRC036-03655 NRC036-03655 T-C--/ý k/- f6C', Di' o;y*

Docket DocketNumberNumber50-346 SQ..346 License Number License Number NPF-3 NPF-3 Serial SerialNumber Number27312731

.Page Page22 If Ifyou youhavehaveany anyquestions, questions, or orrequire require further furtherinformation, information,please pleasecontact contact Mr.

Mr. David H. Lockwood, Manager, Regulatory Affairs, at(419)

David H- Lockwood, Manager, Regulatory Affairs, at (419) 321-8450.

321-8450.

. Very truly

,Very trulyyours, yours,

..~>~~h-l1 t.

RMC/s RMc/s

, Enclosure and Attachments,..

cc:

cc: J.J. E.

E. Dyer, Dyer.Regional RegionalAdministrator, Administrator,NRC NRC Region m Region Ill S.S. P. Sands, DB-1 NRC/NRR Project P. Sands, DB-l NRCINRR Project Manager Manager K.

K. S.S. Zellers, Zellers.DB-1 DB-l Senior SeniorResident Resident Inspector Inspector Utility Radiological Safety Board Utility Radiological Safety Board

((

NRC036-03656 NRC036-03656

Docket Number 50-346 License Number NPF-3 NPF-3 Serial Number 2731 Enclosure Page 11 of 1 RESPONSE TO TO BULLETIN 2001-01 NRC BULLETIN FOR DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1I This letter is submitted pursuant to 10 CFR 50.54(f) 50.54(f) and contains information pursuant to NRC Bulletin 2001-0 1, "Circumferential Cracking of Reactor 200 1-0 I, Reactor Pressure Vessel Head Penetration Penetration Nozzles," for the Davis-Besse Nuclear Power Station, Unit Number 1. 1.

(1) I am Vice President - Nuclear I, Guy G. Campbell, state that (1) Nuclear of the FirstEnergy FirstEnergy Nuclear Nuclear Operating Company, (2) I am duly authorized to execute execute and file this certification on behalf of of the Toledo Edison Company and The Cleveland Electric Dluminating Illuminating Company, and (3) the statements set forth herein are true and correct to the best of my knowledge, information and belief.

BY:~~

By:

GU}tG:C Guy G. C mpbell,Vice4esaent pbell:Vice . ent - Nuclear Affirmed and subscribed subscribed before me this 4th day of September, September, 2001.

2001.

Notary Notary Public, State of Ohio - Nora L. L. Flood expires September 4, commission expires My commission 4, 2002.

2002.

NRC036-03657 NRC036-03657

Docket Number 50-346 License Number NPF-3 Serial Number 2731 I Page 11 of 19 19

Response

Response to NRC Bulletin 2001-01 Davis-Besse Nuclear Power Station 2001-01 for the Davis-Besse Station The following infonnation information is provided in response response to NRC Bulletin 2001-01, 2001-01, "Circumferential "Circumferential Cracking of Reactor Reactor Pressure Vessel Head Penetration Penetration Nozzles,"

Nozzles," for the Davis-Besse Nuclear Power Station Station (DBNPS).

NRC Bulletin Request Item 1.a:

Bulletin Reguest l.a:

The plant-specific plant-specific susceptibility ranking for your plant(s) (including all data used to determine each ranking) using the PWSCC susceptibility model described described in Appendix Appendix B to the MRP-44, Part 2, report.

Response

The DBNPS has been analyzed for susceptibility susceptibility relative to the Oconee Nuclear Nuclear Station, Unit 3 (ONS3) using the Materials Reliability Reliability Program (MRP) time-at-temperature time-at-temperature Primary Primary Water Stress Corrosion Corrosion Cracking (PWSCC) model. The parameters used in this ranking are included in Attachment evaluation showed that it will take the DBNPS 3.1 Effective Full Power . This evaluation Power Years (EFPY) of additional operation from March 1,2001, Years (EFPy) 1, 2001, to reach the same time-at-temperature temperature as ONS3 when leaking nozzles were discovered in March 2001.

March 2001.

The DBNPS falls into the NRC category category of plants within 5 EFPY of ONS3.

NRC Bulletin Request Item 1.b:

Bulletin Reguest A description of the VHP nozzles in your plant(s), including including the number, type, inside and outside diameter, materials of construction, construction, and the minimum distance between between VHP nozzles.

Response

The DBNPS has 69 Control Rod Drive MechanismMechanism (CRDM) nozzles of which 61 are used for CRDMs, 7 are spare, and one is used for the Reactor Pressure Vessel (RPV) head vent piping which extends from the CRDM nozzle and terminates at the top of Steam Generator Generator Number Number 2.

Each CRDM CRDM nozzle is constructed constructed of Inconel Alloy 600 and is attached attached to the RPV head by an Inconel Alloy 182 J-groove I-groove weld. The RPV head is constructed of carbon steel and is internally internally clad with stainless steel.

steel, The material for the nozzles was supplied by two suppliers. B&W B&W Tubular Products supplied supplied material material for 60 nozzles and Huntington Alloys supplied the material for the remaining remaining 9 nozzles. The head arrangement arrangement and requested requested nozzle details are provided in Attachment . .

NRC Bulletin Request Reguest Item 1.c:

I.c:

A A description description of the RPV head insulation type and configuration.

NRC036-03658 NRC036-03658

Docket Number 50-346 License Number NPF-3 NPF-3 Serial Number 2731 I Page 22 of 1919

Response

The DBNPS has metal reflective horizontal vessel head insulation. Metal reflective insulation is is "

used on the exterior of the vessel from the closure flange down to and including the exterior of the bottom head dome. Removable metal reflective insulation panels enclose the top head closure flange and studs. Metal reflective insulation is used on the RPV head. A gap exists between the RPV head and the insulation, the minimum gap being at the dome center of the RPV head where it is approximately 2 inches, and does not impede a qualified visual inspection. This 7749-M-197-2-3 of the general arrangement outline is shown in the attached DBNPS drawing 7749-M-197-:2-3 for the RPV insulation.

NRC Bulletin Request Reuuest Item l.d:

A description of the VHP nozzle and RPV head inspections (type, scope, qualification requirements, and acceptance criteria) that have been performed at your plant(s) in-the requirements, in the past 4 .

years, and the findings. Include a description description of any limitations (insulation or other impediments) to accessibility impediments) accessibility of the bare metal of the RPV head for visual examinations.

Response

The DBNPS has performed inspections within the past four years, during the 11th performed two inspections I VI Refueling Refueling Outage (RFO) in April 1998 and during the 12th 12d RFO in April 2000. The scope of the visual inspection was to inspect the bare metal RPV head area that was accessible through the weep holes to identify any boric acid leaks/deposits. The DBNPS also inspected 100% 100% of Control Control Rod Drive Mechanism Mechanism (CRDM) flanges for leaks in response to Generic Letter 88-05, "Boric Acid Corrosion of Carbon Steel Reactor Pressure Pressure Boundary Boundary Components Components in PWR Plants."

Plants." The *results results of these two recent inspections are described below.

Inspections of the RPV head are performed with the RPV head insulation installed in accordance accordance with DBNPS procedure NG-EN-00324, "Boric procedure NG-EN-00324, "Boric Acid Corrosion Control Control Program,"

Program," which was developed in developed in response respOnse to Generic Letter 88-05. As stated previously, previously, a gap exists exists between between the RPV head and the insulation, the minimum gap being at the dome center center of the RPV RPV head where it is approximately approximately 2 inches, inches, and does not impede visual inspection. The service service structure structure envelopes the envelopes the DBNPS DBNPS RPV RPV head head and has 1818 openings (weep holes) at the bottom bottom through which inspections inspections are performed.

performed. There There are are 69 CRDM nozzles nozzles that penetrate penetrate the RPV head. The metal reflective reflective insulation insulation is located located above the head head and does not interfere interfere with the visual inspection.

The visual visual inspection inspection is performed performed by the use of a small camera. This camera is inserted inserted through the weep weep holes.

• April 1998 1998 Inspection Inspection Results (1 IRFO)

(lIRFO)

This visual inspection This visual inspection showed an uneven uneven layer ofof boric boric acid deposits scattered over the head.

There There were were some some lumps lumps of of boron, boron, with the color color varying varying from brown to white. The The outside diameter diameter ofof the the CRDM CRDM tubes tubes showed white white streaks, streaks, providing providing evidence evidence ofof downward downward flow and attributable attributable to CRDM CRDM flange flange leakage. The head was cleaned cleaned by use use of a manual NRC036-03659 NRC036-03659

Docket Number Number 50-346 50-346 License Number NPF-3NPF-3 Serial Num~r Number 2731 Attachment Page 3 of 19 scrubber scrubber and vacuum through the weepholes.

weepholes. The headhead was videotaped after cleaning for future reference.

reference. .

• April 2000 Inspection Results (12RFO)

(12RFO)

In April 2000, Framatome Nuclear Nuclear Power Services Services performed a 100%100% video inspection of of CRDM flanges above the RPV insulation. Five leaking leaking CRDM flanges were identified identified at F1O, OW, locations FlO, Cl1, F8, and G9 DlO, Cll, G9... The main source of leakage was associated ofleakage associated with the D1O CRDM flange. Positive eviderice 010 evidence (boron deposits on the vertical faces of the CRDM flanges and nozzle) existed existed that drives PS, F8, FlO F10 and Cll Cl 1 had had limited gasket leakage.

leakage. CRDM G9 had boron deposits under the CRDM flange between the flange and insulation, providing confidence that this leakage leakage was associated with flange leakage.

leakage. All five CRDM gaskets were replaced and the D D10 a 1 CRDM flange was machined. Visual inspection of the flanges was performed. Some boric boric acid crystals had accumulated accumulated on the RPV head insulation beneath the leaking beneath leaking flanges. These deposits were cleaned (vacuumed). After cleaning, the area above the insulation was videotaped for future reference.

reference.

Inspection Inspection of the RPV head/nozzles indicated some accumulation of boric acid deposits.

head/nozzles area indicated The boric acid deposits were located located beneath the leaking flanges with clear evidence evidence of of downward downward flow. No visible evidence of nozzle leakage was detected. The RPV head area was cleaned cleaned with demineralized demineralized water to the greatest extent possible possible while maintaining the principles of As-Low-As-Reasonably-Achievable As-Low-As-Reasonably-Achievable (ALARA) regarding regarding the dose. Subsequent Subsequent video video inspection inspection of the cleaned cleaned RPV head areas and nozzles was performed performed for future reference.

reference.

• Subsequent Review of 1998 and 2000 Inspection Inspection Videotapes Videotapes Results Since May May 2001, 200 I, a review review of the 1998 1998 and 2000 2000 inspection videotapes of the RPV head has been performed.

performed. This review was conducted conducted to re-confirm re-confirm the indications of boron leakage experienced experienced at the DBNPS OBNPS were not similar similar to the indications indications seen at ONS and ANO- ANO-l; 1; i.e.,

was not indicative indicative of RPV nozzle nozzle leakage. This review determined that indications indications such as those that would result from RPV head penetration penetration leakage were not evident.

NRC Bulletin Request Item 1.e: I.e:

A description description of the configuration configuration of the missile shield, the CRDM CRDM housings and their support/restraint system, and all components, support/restraint components, structures, and cabling from the top of the RPV head up to the missile shield. Include the elevations of these items relative to the bottom of of the missile shield.

Response

The lower section of the service service structure structure is welded welded to the head. The service service structure then bolts to this lower section. Fan holes are provided provided to allow forced air cooling cooling of CRDMs. Ductwork connected to two remotely remotely mounted, 100 100 percent capacity cooling fans is mounted mounted over the fan holes in the service structure. The lower portion of the serviceservice structure structure is also provided with NRC036-03660 NRC036-03660

Docket Number 50-346 License Number NPF-3 Serial Number Number 2731 Attachment 1 Page 4 of 19 ledges to support the RPV head insulation. The upper portion of the service service structure cylinder cylinder is provided with a monorail accommodate chain hoists that are required monorail to accommodate required for stud tensioner handling. A deck is provided on the service structure to provide a work platform for servicing the CRDMs. This deck also provides the support for the CRDM cooling water manifolds and electrical cables. The deck is composed of individual butted plates with openings to accept accept seismic clamps provided with the CRDMs. These seismic plates provide provide stability for the upper portion of the CRDM. They are field-aligned to the reactor vessel control rod nozzles Additional components that are located above the RVP head and below Additional below the missile shield within within the refueling canal include the RPV head vent line piping, CRDM cabling, cooling cooling water piping piping for CRDM thermal barriers, and miscellaneous miscellaneous electrical power cables.

The elevations for the Reactor Coolant System (RCS), including including the top of the CRD Closure service structure, are shown in the attached Housings at the top of the service Figure 1 and Figure 2. The attached Figure service structure is at elevation 653'0". The missile shield is top of the missile shield over the service comprised of six concrete comprised removable panels, each 31' concrete removable 5" x 6' 6" 31' 5" 6" x 3'. It spans the refueling canal and is supported on both sides by the SteamSteam Generator "D-Ring" "D-Ring" walls.

NRC Bulletin Request Item 2:

If your plant has previously experienced leakage from or cracking in VHP nozzles, experienced either leakage addressees addressees are requested to provide the following information:

information: [a, b, c, d]

Response

leakage from or cracking of its RPV head previously identified either leakage The DBNPS has not previously penetration penetration nozzles.

NRC Bulletin Request Request Item 3.a:

susceptibility ranking for your plant is within 5 EFPY of ONS3, addressees are If the susceptibility requested to provide the following information:

requested inspections (type, scope, qualification

a. your plans for future inspections

a. qualification requirements, acceptance requirements, and acceptance criteria) and the schedule.

Response

The DBNPS plans for future inspections consist of the following:

examination of the RPV head will be performed during 13RFO, which is

1. A qualified visual examination 1.

scheduled for April 2002.

currently scheduled NRC036-03661 N RC036-03661

Docket Number Number 50-346 License Number NPF-3 Serial Number 2731 Attachment Page 5 of 19 Visual examinations have been perfonned performed during each refueling outage and reviewed reviewed by the engineering staff. For the 13RFO, a qualified visual examination engineering examination will be perfonned.

performed. Personnel performing this task will be instructed on the type of unacceptable unacceptable conditions using ONS3 ONS3 as the basis. Inspections Inspections will be perfonned performed in accordance accordance with a procedure procedure developed specifically specifically for these examinations requirements of an ASME VI' examinations that will meet the basic requirements VT-2-2 inspection, and will not be compromised compromised due to any pre-existing pre-existing boric acid crystal deposits. The previous inspection video of the cleaned cleaned head and flanges will be used to help determine determine any unacceptable unacceptable conditions. The RPV head will be cleanedcleaned (as (as necessary) and videotaped prior to return to re-establish a baseline for future inspections.

service to re-establish service acceptance criteria to be used will consist of comparative The acceptance evaluations of any as-found boric comparative evaluations acid crystal crystal deposits to photographs of leaking CRDM CRDM nozzles observed at ONS3 and Arkansas Nuclear One-Unit (ANO-1) and evaluation against any identified One-Unit 1 (ANO-I) identified leaking leaking CRDM nozzle nozzle flanges. The cracks leading to the leak will be characterized supplemental examination characterized by supplemental examination and and the nozzle nozzle will be repaired.

repaired.

Because Because there are significant efforts being being undertaken undertaken by the MRP and the nuclear industry to better better understand this phenomena phenomena and to develop develop optimized optimized inspections inspections methods methods (including (including tooling), mitigation and repair techniques, the foregoing is an interim response response to NRC Bulletin Request Request 3.a reflecting reflecting the current CUlTent plans based on information information currently currently available. The FirstEnergy FirstEnergy Nuclear Operating Operating Company (FENOC) proposes to provide provide a final response to NRC Bulletin Request 3.a by January 29, 2002 2002 (60 days before the start of 13RFO 13RFO scheduled scheduled for the spring of 2002).

2(02). Final plans will be based on the inspection inspection results from other facilities, the ongoing ongoing work of the MRP, MNP, and the advancement advancement of Non-Destructive Non-Destructive Examination Examination (NDE)(NOE) technology technology and development development of remote tooling adequate to perform perfonn effective effective and timely surface surface or volumetric examinations examinations from underneath underneath the RVP head.

A flow chart of the inspection plan is shown in Figure 3. Details of the inspection plan will be developed prior to the 13RFO.

2. Qualified Qualified visual examinations examinations will continue continue to be performed at subsequent refuelingrefueling outages.

The DBNPS will continue to perfonnperform qualified qualified visual examinations examinations of the RPV head for evidence of leaking CRDM nozzles at subsequent refueling refueling outages. The visual examination examination procedure will be updated, as required, required. to include industry industry experience.

NRC Bulletin ReMuest Reguest Item 3.b:

concluding that the inspections Your basis for concluding inspections identified identified in 3.a. will assure that regulatory requirements areare met (see Applicable Applicable Regulatory Regulatory Requirements section). Include the following specific information in this discussion:

specific infonnation (1) If your future inspection plans do not include performing performing inspections before before December December 31, 2001,

31. 2001, provide your basis for concluding that the regulatory regulatory requirements requirements discussed in in NRC036-03662 N RC036-03662

50-346 Docket Number 50-346 License Number NPF-3 Number 2731 Serial Number Attachment 1 Page 6 of 19 19 the Applicable Regulatory Requirements section will continue to be met until the Regulatory Requirements inspections performed.

inspections are perfonned.

inspections, discuss the corrective inspection plans include only visual inspections, (2) If your future inspection actions that will be taken, including alternative inspection methods (for example, volumetric examination), if leakage is detected.

volumetric examination),

Response

The DBNPS is similar in design to ONS3 and ANO-l, ANO-1, which have demonstrated demonstrated an ability to CRDM nozzles by visual inspection for boric acid crystal deposits. This has identify leaking CRDM been examination of additional been demonstrated at these units by examination additional non-leaking non-leaking nozzles for signs of of cracking. In each of the twenty-six nozzles, the results did not find any signs of significant significant cracking, thereby providing the necessary necessary confidence confidence that leaking CRDMCRDM nozzles can be found by visual inspection. The DBNPS fabrication records were reviewed determine how CRDM reviewed to determine bores were machined and how CRDM nozzles were installed. CRDM nozzles were installed installed in the RPV closure head with a designed 0.0005 inches to 0.0015 inches of diametral diametral interference interference (documented in "Safety (documented B&W-Design Reactor Vessel Head Control Rod Drive "Safety Evaluation for B&W-Design Mechanism BAW-10190P, dated May 1993). The CRDM nozzle shaft Cracking," BAW-I0190P, Mechanism Nozzle Cracking,"

diameter is custom ground to 0.001 inches greater than the final diameter of the associated CRDM bore with a 32AA 32AA finish. A general description of the CRDM bores machining machining is as follows:

• a Rough machine CRDM bores (Note: DBNPS RPV head penetrations penetrations were not counterbored.)

c,ounterbored.)

treatment of RV closure head

• Final heat treatment

• • Finish machine CRDM bores to a 250 finish A general description of the CRDM nozzle installation is as follows:

• • Cool CRDM nozzles in liquid nitrogen to -140°F-140°F minimum minimum

• • Install CRDM nozzle in specified specified location

• • Allow CRDM nozzle to warm to 70°F During the final Quality Assurance inspection, CRDM bores were inspected inspected for final top and and bottom bore diameter individual CRDM nozzle shaft custom grinding verticality. After individual diameter and verticality. grinding to approximately inches greater in diameter than the final CRDM approximately 0.001 inches CRDM bore diameter, CRDM measured at both the top and the bottom of the custom ground length.

nozzle shafts were also measured CRDM nozzle shafts are longer than CRDM bores are deep. Thus, CRDM nozzle shaft diameter measurements do not directly line up with CRDM bore diameter measurements measurements, although in the diameter measurements, counterbores.

case of the DBNPS these locations should be fairly close because of the lack of counterbores.

Therefore, the resulting resulting top and bottom dimensional fits are considered considered approximate. The values for the DBNPS RPV head are calculated to range from a maximum interference interference fit of 0.0021 inches to a gap of 0.0010 inches.

NRC036-03663 NRC036-03663

Docket Number Docket Number 50-346 License Number License Number NPF-3 Serial Number Serial Number 2731 Attachment Page Page 7 ofof 19 In 1993, 1993, the the B&WOG B&WOG performed perfonned a safety evaluation evaluation for CRDM CRDM nozzle nozzle cracking cracking (reference:

(reference:

previously cited BAW-10190P).

previously BAW-10190P). In this evaluation, a 3D finite element element model model of all major major components of a hillside components hillside CRDM CRDM nozzle-to-head structure was constructed. The nozzle-ta-head welded structure B&WOG calculation includes the B&WOG calculation 0.010 inch diametric the maximum 0.010 counterbore at the top and diametric counterbore and bottom locations (typical (typical for most B&WOG B&WOG plant designs),

designs), which which tends tends to increase increase the the stresses in the nozzle bounding for the DBNPS. During operation, an interference nozzle and is bounding interference fit is calculated become a gap release to become to release temperature and gap due to temperature and pressure provides a leak path pressure dilation, which provides for a through-wall through-wall crack crack that that allows allows detection inspection. The B&WOG detection by visual inspection. B&WOG calculation calculation assumes assumes a nominal nominal 0.001 inch inch interference interference fit, which which will will open to a maximum maximum: gap gap of 0.0033 0.0033 inches during operation.

As noted earlier, leakage leakage from this gap has been demonstrated demonstrated at both ONS and ANO-!, ANO- 1, for which interference interference fits of up to 0.0014 0.00 14 inches have been calculated have been calculated from the final QA inspection data (as documented documented in MRP-44, MRP-44, Part Part 2). Figure Figure 4 provides a graphical graphical representation representation of these these data. The largest interference interference fit at the DBNPS occurs on nozzle number number 50 which, as stated previously, has been calculated calculated at 0.0021 inches inches at the top. This same nozzle nozzle also has an an interference interference fit of 0.0010 0.0010 inches at the bottom. Thus, the 0.0033 inch gap during operation would be somewhat somewhat less for the DBNPS, DBNPS, assuming assuming the 0.0021 inch interference interference fit (instead of thethe nominal nominal 0.001 inch). This gap would would still be expected to provide provide a leak path to the top of the RPV head in the event event of a cracked cracked CRDM CRDM nozzle nozzle or J-groove weld. The DBNPS DBNPS has not not observed observed any leakage from these paths during its past inspection activities.

The DBNPS plans to perform perfonn inspections of the RPV head and CRDM nozzles as recommended recommended by MRP-48. The inspections inspections will consist of qualified visual inspections inspections of the top RPV head bare metal metal surface at the 13RFO 13RFO scheduled for the spring of 2002. If any leaks are detected, the source will be determined, determined, the cracks leading to the leak will be characterized characterized by supplemental examination and the nozzle will be repaired.

significant efforts being As stated previously, because there are significant undertaken by the MRP and the being undertaken nuclear industry to better understand understand this phenomena phenomena and to develop optimized inspections methods (including tooling), mitigation and repair techniques, the foregoing is an interim response reflecting the current plans based on information information currently available. The FENOe FENOC proposes to provide a final response by January January 29, 2002 (60 days before the start of 13RFO 13RFO 2002). Final plans will be based on the inspection results from other scheduled for the spring of 2(02).

facilities, the ongoing work of the MRP, and the advancement of NDE technology technology and development of remote tooling adequate development adequate to perform effective and timely surface or volumetric volumetric underneath the RVP head.

examinations from underneath The Applicable Regulatory Requirements section of the Bulletin lists the following regulatory regulatory requirements and plant commitments as providing the basis for the Bulletin assessment:

• • Appendix A to 10 CFR 50, "General Design Criteria for Nuclear Power Plants" Boundary" Criterion 14 - "Reactor Coolant Pressure Boundary" NRC036-03664 NRC036-03664

Docket Docket Number 50-346 License Number NPF-3 License NPF-3 .

Serial Serial Number 2731 Attachment Attachment 1 Page 8 of 19 19 "Fracture Prevention of Reactor Coolant Criterion 31 - "Fracture Boundary," and Coolant Boundary," and Criterion 32 - "Inspection Criterion Reactor Coolant Pressure Boundary" "Inspection of Reactor Boundary"

• • Plant Technical Technical Specifications

• • 10 CFR 50.55a, Codes and Standards, which incorporates incorporates by reference Section XI, "Rules "Rules for Inservice Inspection of Nuclear Power Plant Components, of the ASME Boiler and Pressure Vessel Code"

• • Appendix B of 10 CFR CPR 50, "Quality "Quality Assurance Assurance Criteria for Nuclear Nuclear Power Plants and Fuel Reprocessing Plants," Criteria V, ~'Instructions, Reprocessing Procedures, and Drawings;"

"Instructions, Procedures, "Control of Drawings;" IX, "Control Special Processes;" "Corrective Actions" Processes;" and XVI, "Corrective Actions" The following addresses each of these criteria and demonstrates that the criteria will be met for The following addresses each of these criteria and demonstrates that the criteria will be met for the DBNPS until the inspections inspections are performed.

Design Requirements:

Requirements: 10 CFR AI?pendix A - General CPR 50. Appendix General Design Requirements Reguirements The Bulletin states:

"The

The applicable ODC [General Design Criteria] include GDC [General include GDC ODC 14, GDC 31, and GDC 32. GDC ODe 31, 14 specifies that the reactor coolant pressure boundary boundary (RCPB) have an extremely low abnormal leakage, of rapidly propagating failure, and of gross rupture; the probability of abnormal presence of cracked cracked and leaking nozzles is not consistent with this GDC. GDC 31 specifies leaking VHP nozzles propagating fracture of the RCPB be minimized; the presence of that the probability of rapidly propagating cracked and leaking VHP nozzles nozzles is not consistent with this GDC. GDC 32 specifies specifies that components which are part of the RCPB have the capability components capability of being periodically periodically inspected to practices that do not permit reliable assess their structural and leaktight integrity; inspection practices cracking are not consistent with this GDC."

detection of VHP nozzle cracking These referenced referenced criteria state the following:

• Criterion Criterion 1414 -- Reactor Reactor Coolant Pressure Boundary reactorcoolant "The reactor coolant pressure pressure boundary boundary shall shall be designed, designed,fabricated, fabricated, erected and and tested so probability abnormal leakage, rapidly propagating failure, as to have an extremely low probability of abnormal leakage, of rapidlypropagating failure, and of gross rupture."

gross rupture. "

• • Criterion Criterion 3131 -- Fracture Fracture Prevention Prevention of Reactor Coolant Pressure Pressure Boundary "The reactor coolantpressure reactor coolant boundary shall pressure boundary designed with sufficient margin shall be designed assure margin to assure that when stressed that under operating, stressed under operating,maintenance, maintenance,testing, testing, andpostulated accident conditions postulated accident conditions (1)

( 1) the boundary behaves in a non-brittle boundarybehaves non-brittle manner, probability of rapidly and (2) the probability manner, and rapidly propagatingfracture propagating minimized. The design fracture is minimized. design shall considerationof service shall reflect consideration service temperatures temperatures and and other conditions conditions of the boundary materialunder operating, boundary material maintenance, operating. maintenance, NRC036-03665 NRC036-03665

Docket Number Number 50-346 License Number NPF-3NPF-3 Serial Number 2731 Attachment 1 Attachment Page 9 of 19 testing and andpostulated postulatedaccident accidentconditions conditionsand and the uncertainties uncertaintiesin determining material determining (1) material properties, properties, (2) the effects of irradiation materialproperties, irradiation on material properties,(3) residual, steady state and residual, steady transient transient stresses, and (4) stresses, and (4) size of offlaws.

flaws. ""

0* Criterion 32 -- Inspection Criterion 32 Inspection of Reactor Reactor Coolant Pressure Boundary "Components which are "Components arepart reactor coolantpressure part of the reactor boundary shall be designed pressure boundary designed to periodic inspection permit (1) periodic inspection and important areas and testing of important andfeatures areas and assess their features to assess aiul (2) integrity,and and leak tight integrity, structural and structural (2) an appropriate appropriatematerial program for surveillance program materialsurveillance the reactor pressurevessel."

reactorpressure vessel. "

demonstrated that the design of the reactor coolant During initial licensing of the DBNPS it was demonstrated requirements in place at that time. The ODe pressure boundary met the requirements GDC included in Appendix Appendix A to 10 CFR SO effective until May 21, 50 did not become effective 21, 1971. construction permit 1971. The construction pennit for the DBNPS was issued prior to May May 21, 1971; consequently, 21.1971; consequently, the DBNPS was not subject to the requirements (reference:

GDC requirements SECY-92-223; 9/18/92). However, the following demonstrates (reference: SECY-92-223; criteria for the RPV head nozzles.

compliance with the design criteria compliance

• • Pressurized Pressurized water reactors licensed licensed both before and after issuance of Appendix A to Part SO 50 (1971) complied with these criteria in part by: 1)

(1971) 1) selecting Alloy 600, and other austenitic materials with excellent excellent corrosion resistance resistance and extremely extremely high fracture toughness, for reactor reactor coolant pressure boundary materials, and 2) following ASME Codes and Standards and other applicable requirements erection, and testing of the pressure requirements for fabrication, erection, pressure boundary parts. NRC reviews 'of of operating license submittals subsequent subsequent to issuance of of Appendix A included evaluating evaluating designs for compliance compliance with the General Criteria.

General Design Criteria

• • Although stress corrosion cracking of primary coolant corrosion cracking coolant system penetrations was not originally originally anticipated during plant design, it has occurred in the RPV top head nozzles nozzles at some plants.

The suitability of the originally selected materials has been confirmed. The robustness of the design has been demonstrated by the small amounts of the leakage that has occurred and by cracks in Alloy 600 reactor coolant pressure boundary the fact that none of the cracks boundary materials has propagated or resulted in catastrophic rapidly propagated catastrophic failure or gross rupture. Given the inherently inherently high fracture toughness tolerance of the Alloy 600 material toughness and flaw tolerance material there is indeed an extremely extremely low probability of a rapidly propagating propagating failure and gross rupture. It should be noted that the originally originally proposed Appendix Appendix A (July 1967)1967) was written in terms of extremely extremely low probability probability of gross rupture or significant significant leakage throughout the design life.

• • Utilizing the conservative conservative time-at-temperature time-at-temperature ranking model of MRP-44,MRP-44, the operating time before Davis-Besse would reach an equivalent equivalent degradation degradation time as ONS-3 is at least 3.1 EFPY.

• • An updated updated safety assessment was performed performed by Framatome-ANP Framatome-ANP in April 2001 to address the CRDM nozzle cracking observed at ONS-1, observed ONS-l, ONS-3, and ANO-1.

ONS-3, ANO-1. Flaw growth growth calculations performed, using the modified Peter Scott crack growth calculations were performed, growth equation equation and assuming an initial flaw length of 180 18000 around the nozzle, which indicate indicate that it would take approximately 4 years for a through-wall flaw to grow another 25% around the approximately circumference. This circumference. This remaining remaining ligament, ligament, which would be be 25%

25% of the the original circumference, origina] circumference, NRC036-03666 NRC036-03666

Docket Number 50-346 License Number NPF-3 Serial Number Serial Number 2731 Attachment 1 Attachment Page 10 of 19 would still be sufficient to preclude preclude gross net-section failure (nozzle ejection). This ligament ligament satisfies satisfies primary stress limits using a safety factor of 3. 3.

• The revised The revised Framatome Framatome ANP safety safety assessment April 2001 also concluded that assessment of Apri12001 simultaneous simultaneous multiple CRDM nozzles nozzleS will not fail and that the failure of a single CRDM nozzle is bounded by both theLOCAthe LOCA and non-LOCA plant analyses already completed completed to support current current plant operation.

• " MRP-44, Appendix MRP-44, Appendix C C describes describes the accident sequence sequence analyses already in place using the DBNPS Emergency Operating Operating Procedures (EOPs). The existing EOPs provide adequate directions to mitigate any transient that would occur occur should there be a failure of a CRDM nozzle.

• " All evidence All evidence toto date date suggests that it suggests that it will require several several years for the material to degrade to the point that total failure of the component component could occur. During that time, if a crack should form, leakage of primary coolant on the RPV head can be identified through routine visual inspection of the bare RPV head. The component component can then be repaired repaired and returned to

,service without jeopardizing jeopardizing the health and safety of the public.

Therefore, the requirements requirements established established for design, fracture toughness, toughness, and inspectability GDC inspectability in ODC 14, 3 1, and 32, respectively, were satisfied during the initial licensing 14,31, licensing review review for the DBNPS, DBNPS, and continue to be satisfied during operation even in the presence presence of a potential potential forstress corrosion cracking of the RPV head penetrationpenetration nozzles.

Operating Operating Requirement:

Requirement: 10 CFR 50.36 - Technical Specifications Specifications The Bulletin states:

"Plant technical specifications pertain technical specifications pertain to the issue of VHP nozzle crackingcracking insofar as they require no through-wall through-wall reactor coolant system leakage." leakage."

10CFR IOCFR 50.36 contains contains requirements requirements for Plant Technical Technical Specifications. Paragraphs Paragraphs 2 and 3 of 10 CFR 50.36 are particularly particularly relevant:

• 1IOCFR 50.36(c)(2) 10CFR 50.3 6(c)(2) Limiting Conditions Conditions for Operation Operation

"(i)

"( i) Limiting conditions conditionsforfor operation operation areare the lowest functional capability orperformance functional capability performance equipment required levels of equipment requiredfor safe operation operation of the facility.

facility. When a limiting conditionfor limiting condition operationof a nuclear operation nuclear reactor met, the licensee reactor is not met, licensee shall shall shut down the reactor orfollow reactor or follow any remedial remedial action actionpermitted pennitted by the technical specifications until technical specifications until the condition condition can be met.

(ii)A technical (ii) technical specification specification limiting limiting condition condition for operation operationof aa nuclear reactormust be nuclear reactor established established for each each item meeting one or more of the following criteria:

criteria:

(C) Criterion (C) Criterion3: A structure, structure, system, or component component that is partpart of the primary success path primary success path and and which functions or actuates to mitigate a design basis accident functions actuates mitigate design basis accident transient that or transient that either assumes the failure assumes failure of or presents presents a challenge challenge to the integrity integrity of a fission fission product product barrier.

barrier.

NRC036-03667 N RC036-03667

Docket Docket Number Number 50-346 License Number License Number NPF-3 Serial Serial Number Number 2731 Attachment Page Page 1111 of 19 (D) Criterion4: A structure, system, or component which operatingexperience or (D) Criterion 4: A structure, system, or component which operating experience or probabilisticrisk assessment probabilistic significant to public health andsafety. "

assessment has shown to be significant

• 10 CFR 50.36(c)(3) Surveillance Requirements 50.36(c)(3) Surveillance Requirements "Surveillance "Surveillancerequirements relating to test, requir.ements relating requirementsare requirements test, calibration, inspection to calibration, or inspection to assure that the necessary quality of systems and components is maintained, assure thatfacility maintained, that facility limits, and that the limiting conditions operation will be within safety limits.

operation operation will be conditionsfor operation met."

The reactor reactor coolant coolant pressure pressure boundary provides one of the critical boundary provides critical barriers barriers that guard against thethe uncontrolled uncontrolled release radioactivity. Therefore, release of radioactivity. Therefore, the DBNPS Technical Technical Specification Specification 3.4.6.2 3.4.6.2 includes a requirement and associated action statements addressing includes reactor coolant pressure addressing reactor boundary leakage. The limits for reactor boundary leakage. reactor coolant pressure pressure boundary boundary leakage stated in terms leakage are stated amount of leakage, of the amount leakage, e.g.,

e.g .* 1 gallon per minute for unidentified unidentified leakage; <10

~1O gpm for identified leakage; and no reactor coolant system pressure boundary leakage.

identified leakage.

Leaks from Alloy 600 600 RVP head penetrations penetrations due to PWSCC have have been well below below the sensitivity of on-line leakage detection systems. Plants leakage detection evaluated this condition and have Plants have evaluated determined appropriate inspections are bare-metal visual inspections for boric acid determined that the appropriate acid deposits during plant shutdowns. If leakage leakage or unacceptable indications are found, the defect must be repaired before the plant goes back on line. If repaired before If through-wall through-wall boundary leaks of the CRDM nozzles increase increase to the point where they are detected by the on-line leak detection Specification's specified systems, then the leak must be evaluated per the Technical Specification's systems. acceptance specified acceptance Specification's required actions taken.

criteria and the Technical Specification's Requirements: 10 CFR Inspection Requirements: CFR. 50.55a and ASME Section XI The Bulletin states:

"NRC regulations CFR 50.55a state that ASME Class 11 components regulations at 10 CFR. components (which include VHP nozzles) must meet the requirements of Section XI of the ASME Boiler and Pressure Vessel 1

Code. Table IW 1WA-2500-1 [MWB-2500-1 1]

A-25OO-1 [IWB-25OO-1 ] of Section XI of the ASME Code provides examination requirements for VHP nozzles and references IWB-3522 for acceptance acceptance standards.

conditions requiring correction IWB-3522.1(c) and (d) specify that conditions IWB-3522.1(c) correction include the detection of leakage from insulated components and discoloration or accumulated residues on the surfaces of components, insulation, or floor areas which may reveal evidence of borated water leakage, with components, leakage defined as "the through-wall leakage that penetrates the pressure retaining membrane. membrane.""

11An erratum appears to exist in the Bulletin. Table IW A-2500-1 is cited, but does not exist. It IWA-2500-1 IWB-2500-1.

the citation should have been IWB-2500-1.

appears .the NRC036-03668 NRC036-03668

Docket Docket Number Number 50-346 50-346 License License Number NumberNPF-3 NPF-3 Serial Serial Number Number 2731 2731 Attachment Attachment I1 Page Page 12 12 of of 1919 "Therefore, "Therefore, 10 10 CFR CFR 50.55a, SO.SSa, through through itsits reference reference to to the the ASME ASME Code, Code, does does not not permit permit through-through-wall cracking of VHP wall cracking of VHP nozzles.>> nozzles."

"For "For through-wall through-wall leakageleakage identified identified byby visual visual examinations examinations in in accordance accordance with with the the ASME ASME

Code, Code, acceptance acceptance standards standards forfor the the identified identified degradation degradation are are provided provided in in IWB-3142.

IWB-3142.

Specifically, Specifically. supplemental supplemental examination examination (by (by surface surface oror volumetric volumetric examination),

examination), corrective corrective measures measures or or repairs, repairs. analytical analytical evaluation, evaluation, and and replacement replacement provideprovide methods methods for for determining determining the the acceptability of degraded components."

acceptability of degraded components."

10 10 CFR CFR 50.55a SO.SSa requires requires that that inservice inservice inspection inspection and and testing testing be be performed performed per per the the requirements requirements of the ASME Boiler and Pressure Vessel Code, Section XI, "Inservice of the ASME Boiler and Pressure Vessel Code, Section Xl, "In service Inspection of Nuclear Inspection of Nuclear Plant Components."

Plant Components." Section Section XIXIcontains contains applicable applicable rules rules for for examination, examination. evaluation evaluation and and repair repair of of code code class class components, components. including including thethe reactor reactor coolant coolant pressure pressure boundary.

boundary.

The The DBNPS DBNPS performsperforms visual visual inspections inspections for for evidence evidence of of leakage leakage by by examining examining the the RPV RPV headhead surface surface andand the the CRDM CRDM flanges flanges per per the the requirements requirements of of NRC NRC Generic Generic Letter Letter 88-05, 88-05, "Boric "Boric Acid Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants." If pressure in PWR Plants." If pressure boundary boundary leakage leakage isis suspected, suspected, supplemental supplemental examinations examinations of ofthethe affected affected CRDMCRDM nozzlenozzle willwill be be performed performed to to characterize characterize the the integrity integrity ofof the the nozzle.

nozzle. Some Some plants plants have have conducted conducted inspections inspections beyond beyond those those required required by by Section Section XI XI and and NRC NRC Generic Generic Letter Letter 88-05.

88-05. These These inspections inspections have have included visual examinations of 100% of the bare metal surfaces included visual examinations of 100% of the bare metal swfaces of the RPV head; eddy of the RPV head; eddy current current and and liquid liquid penetrant penetrant surface surface examinations; examinations; and and supplemental supplemental examinations examinations of of the the nozzles.

nozzles. TheseThese supplemental supplemental inspections inspections coupled coupled withwith the the evaluations evaluations of of cracking cracking thatthat has has been been found found areare considered considered to to have have provided provided aa defense-in-depth defense-in-depth approach approach for for investigating investigating and and resolving resolving this this issue.

issue.

The The acceptance acceptance standards standards areare as as detailed detailed in in Technical Technical Specifications Specifications for for pressure pressure boundary boundary leakage leakage since since the the program program under under Generic Generic Letter Letter 88-05 88-05 is is not not aa Code-required Code-required inspection inspection program.

program.

Flaws Flaws identified identified by by supplemental supplemental methods methods willwill be be evaluated evaluated in in accordance accordance with with the the flaw flaw evaluation evaluation rules rules for for piping piping contained contained in in Section Section XI XI ofofthe the ASME ASME Code. Code. AnyAny flaw flaw notnot meeting meeting the requirements for the intended service period would be repaired the requirements for the intended service period would be repaired prior to returning it to prior to returning it to service.

service.

Repairs Repairs to to RPV RPV headhead nozzles nozzles will will be be performed performed in in accordance accordance with with Section Section XI XI requirements, requirements.

NRC-approved ASME Code Case requirements, or an alternative NRC-approved ASME Code Case requirements. or an alternative repair or replacement method repair or replacement method approved approved by by the the NRC.

NRC.

The The DBNPS DBNPS complies complies withwith these these ASME ASME Code Code requirements requirements throughthrough implementation implementation of of the the Inservice Inspection Program. In addition, additional Inservice Inspection Program. In addition. additional inspections are inspections are conducted conducted in in accordance accordance with with the the program program developed developed to to meet meet Generic Generic Letter Letter 88-05.

88-05. IfIf aa VT-2 VT-2 or or qualified qualified visual visual examination detects the cracks or leakage in the CRDM nozzles, examination detects the cracks or leakage in the CRDM nozzles. corrective actions will corrective actions will be be performed in accordance with the DBNPS corrective performed in accordance with the DBNPS corrective action program. No action program. No newnew plant plant actions actions are are necessary necessary to to satisfy satisfy the the regulatory regulatory criteria.

criteria.

NRC036-03669 NRC036-03669

Docket Number Number 50-346 License Number NPF-3 Serial Number 2731 Attachment Page 13 of 19 19 Assurance Requirements: 10 CPR.

Quality Assurance Ouality CFR 50. Appendix B The Bulletin Bulletin states:

"Criterion "Criterion IX of Appendix B to 10 CFR Part 50 SO states that special processes, processes, including including nondestructive nondestructive testing, shall be controlled and accomplished accomplished by qualified personnel using qualified procedures accordance with applicable procedures in accordance specifications, criteria, and applicable codes, standards, specifications, special requirements. Within the context of providing assurance other special assurance of the structural integrity of VHP nozzles, special requirements requirements for visual examination would generally require require the use of a qualified visual examination examination method. Such a method is one that a plant-specific plant-specific analysis has demonstrated demonstrated will result in sufficient leakage leakage to the RPV head surface for a through-wall through-wall crack in a VHP nozzle, and that the resultant leakage provides a detectable detectable deposit on the RPV head. The analysis would have to consider, for example, the as-built configuration of the VHPs and the configuration of capability capability to reliably detect and accurately characterize characterize the source of the leakage, considering considering the presence presence of insulation, preexisting preexisting deposits on the RPV head, and other factors that could could interfere interfere with the detection of leakage. special requirements leakage. Similarly, special requirements for volumetric volumetric examination would generally require examination require the use of a qualified volumetric examination method, for example, one that has a demonstrated example, capability to reliably detect cracking on the OD of the demonstrated capability VHP nozzle above the JJ-groove

-groove weld."

The design shrink fit of the CRDM nozzles at the DBNPS is similar to the design shrink fit of the ONS units indicating that through wall cracking of the nozzles of the magnitude magnitude seen at ONS should produce visually produce visually detectable evidence evidence of leakage leakage on the RPV head. The qualified qualified visual inspection and the personnel involved in the evaluation of the results will be VT-2 qualified qual,ified and familiar with the anticipated type of indication indication that any leakage would cause. Any other NDE equipment that may be required is presently techniques and associated equipment presently being developed andand should be qualified for the DBNPS 13RFO in the spring of 2002.

The Bulletin further states:

"Criterion V of Appendix "Criterion Appendix B to 10 CFR Part 50 states that activities affecting quality shall be prescribed by documented instructions, instructions, procedures, procedures, or drawings, of a type appropriate appropriate to the circumstances circumstances and shall be accomplished in accordance accomplished accordance with these instructions, procedures, procedures, or drawings. Criterion V further states that instructions, procedures, procedures, or drawings shall include appropriate appropriate quantitative quantitative or qualitative acceptance acceptance criteria for determining determining that important activities have been satisfactorily satisfactorily accomplished. Visual and .volumetric examinations of VHP nozzles are nozzles activities that are activities that should be documented requirements."

documented in accordance with these requirements."

The efforts efforts undertaken undertaken to inspect, evaluate, and /or lor repair the DBNPS RPV head penetrations penetrations will be conducted will be conducted andand documented documented in in accordance accordance with with procedures procedures which which comply with the FENOC FENOC Quality Assurance Assurance Program Program and Criterion V of 10 CFR 50, Appendix B.

The final criterion criterion cited by the Bulletin is stated as follows:

NRC036-03670 NRC036-03670

Docket Number 50-346 License Number NPF-3 NPF-3 Serial Number 2731 1 Page 14 of 19 19 "Criterion XVI of Appendix B "Criterion B to 10 10 CFR Part 50 states that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. For significant conditions adverse to quality, the measures taken shall include root cause determination determination and corrective action to preclude repetition of the adverse conditions. For cracking of VHP nozzles, determination is important to understanding the nature of the degradation present the root cause determination present and the required actions to mitigate futu~ future cracking. These actions could include proactive inspections and repair of degraded VHP nozzles."

In addressing Criterion XVI, there are two important attributes pertinent to RPV CRDM nozzles In cracking.

First, Criterion XVI states "Measures shall be established to assure that conditions adverse to quality...

qUality ... are promptly identified and corrected."corrected." This criterion is partially met by the DBNPS's DBNPS's awareness of awareness of industry industry experience, experience, and has been implemented in this manner in the DBNPS corrective action program whereby industry experience experience is evaluated for applicability applicability to DBNPS and the applicable corrective actions, as needed are determined. This is consistent with the communication process, implemented NRC's generic communication implemented by Information Notices, which reports industry experience, but does not require a response to NRC. Licensees are expected to evaluate the applicability of the information contained in the Information Information Notice and document a specific assessment for possible NRC review.

Criterion XVI provides the objectives objectives and goals of the corrective action program, but leaves to the licensee the licensee the responsibility responsibility for determining determining the specific process to accomplish accomplish these objectives objectives and goals. With regard to the Bulletin response, Criterion XVI does not provide specific guidance as to what is an appropriate response, but rather, the licensee licensee is responsible for determining actions necessary determining necessary to maintain public health and safety. In this particular particular instance, the licensee licensee must justify justify its actions for addressing the PWSCC of RPV head head nozzles.

Furthermore, the regulatory Furthermore, the regulatory criteria of criteria of 10 CFR 50.109(a)(7) 10 CFR 50.109(a)(7) provides supporting evidence evidence when when it states states""...if

.. .if there are are two or more ways to achieve achieve compliance..

compliance ....then then ordinarily ordinarily the applicant applicant or licensee is free to choose the way which which best suits its purposes."

purposes." ."

The second The second attribute attribute of Criterion Criterion XVI stated is "In "In the case of significant conditions conditions adverse to quality, the measures shall quality, the measures shall assure assure that the cause of the condition is determined condition determined and corrective corrective action taken action taken to preclude repetition."

to preclude repetition." The Bulletin suggests that for RPV head nozzle cracking, cracking, the the root root cause determination is important to understanding cause determination understanding the nature of the degradation degradation and the required actions to required actions to mitigate future cracking. As part of the DBNPS corrective corrective action program, determination of the cause determination of the cause of.the of the PWSCC PWSCC in in the RPV RPV head nozzles, nozzles, either either through the DBNPS's DBNPS's efforts efforts oror asas part part ofof an an industry industry effort, effort, would would be be performed, performed, if if cracks cracks are are detected.

detected.

In summary, the In summary, the integrated integrated industry industry approach approach to inspection, inspection, monitoring, cause cause determination, determination, and and resolution resolution of of the the identified identified CRDM CRDM nozzlenozzle cracking cracking is in compliance compliance with the performance-performance-based objectives based objectives of 10 CFR 50, Appendix B.

B.

N RC036-03671 NRC036-03671

Docket Number 50-34650-346 License Number Number NPF-3 Serial Number 2731 Attachment Page 15 of 19 NRC Bulletin Reauest Request Item 4:

If the susceptibility ranking for your plant is greater greater than 5 EFPY and less than 30 EFPY of ONS3, ONS3, addressees are requested to provide the following information: [a and b]

Response

This request request does not apply to the DBNPS because susceptibility, ranking because the DBNPS susceptibility ranking is within within 3.1 EFPY of ONS3. I Bulletin Request NRC Bulletin Reauest 5:

requested to provide the following information Addressees are requested Addressees information within within 30 days after plant plant restart following the next refueling refueling outage:

a. a description of the extent of VHP nozzle leakage leakage and cracking detected at your plant, cracking detected including the number, location, size, and nature nature of each each crack detected; detected;

b. if cracking cracking is identified, a description of the inspections inspections (type, scope, qualification qualification requirements, and acceptance acceptance criteria),

criteria), repairs, and other corrective actions you have taken to satisfy applicable regulatory requirements. This information is requested requested only if if there are any changes changes from prior information infonnation submitted in accordance accordance with this bulletin.

Response

The DBNPS will provide the NRC with the following information within 30 days after plant th restart following the 13 13"h scheduled to begin in the spring of 2002:

RFO scheduled

a. A description of the extent of RPV head nozzle leakage leakage and cracking. This information information will include the number, location, size and nature nature of each crack detected.

b. A description description of the inspections (type, scope, qualification requirements, scope, qualification acceptance requirements, and acceptance criteria),

criteria), repairs and other corrective corrective actions taken to satisfy applicable applicable regulatory requirements.

NRC036-03672 NRC036-03672

Docket Number 50-346 Number NPF-3 License Number NPF-3 Serial Number 2731 1 19 Page 16 of 19 NISSILE SHIELD 653' MISSILE SHIELD 653'

........................ j

............................... C~*. ~ ".~~~~~". ~~ ~ ~ '.~ ~ ~". ~~ ~~.......

~~ .. ".. ,.,., PAQ'",

6,

. TOP OF CRD CRO CLOSURE HOUSINGS HOUSINGS 603' -61/2' 603' -6lfz" E

FLIING£ RV FLANGE 578' -0' 563' -6118" TOP OF CORE REACTOR vtSS£L.

Figure 1 Figure NRC03.6-03673 NRC036-03673

Docket Number Number 50-346 50-346 License Number NPF-3 Serial Number 2731 Attachment 1 Attachment Page 17 of 19 INSULATION RV HEAD INSULATION STRUCTURE SERVICE STRUCTURE CRDM SUPPORT SUPPORT STEEL -

18 ACCESS QPENINGS OPENINGS 2" MIN GAP BETWEEN -

2" "MOUSE-HOLES" AT "MOUSE-HOLES" AT INSULATION INSULATION AND TOP TOP DAVIS BESSE OF RV RV HEAD HEAD I I

Figure 2.

Side View Schematic of Davis-Besse Reactor Vessel Head, Head, CRDM Nozzles, and Insulation.

NRC036-03674 NRC036-03674

,:t.!~ 'I" rr.;1,1~ll~';.~j*,.~~'l,1"" j~ ,~,; ~ ~ ~

Docket Number Docket Number 50-346 50-346

.,.1.".****!,~~*L:.,j."l*'.,f'11 ')~,1~1JI,~';*,

License License Number Number NPF-3 l.~~\\' lJU:~:J~L - - Serial Number Serial Number 2731

\\/~~ltl:.~! i.h~'1~HS\~I:tl-'

Attachment 11 Attachment Page Page 18 18 of 19 I M I acid No Boric acid Boric acid acid deposit detected detected deposit deposit detecteddetected Figure 3, CRDM Inspection Inspection I

I Plan Plan Flow Chart Determine Detennine Source Source I

Leaking Leaking uldn't Couldn't Cc CRDM Flange(s) Nozzle(s) Deterrn ne Source Determine Repair Leak Inspect Nozzle(s)

Nozzle(s) 1 I

I Axial Circumferential Crack Crack I

Characterize Flaws Characterize Flaws I

Evaluate Need Evaluate Need for Repair for Repair I T I- I I I Extent of Condition Repair Repair Extent of Condition Inspection Inspection Clean the Head NRC036-03675 N RC036-0367 5

Docket Docket Number Number 50-34650-346 Number NPF-3 License Number License Number 2731 Serial Number Serial Attachment Page 19 Page 19 of 19 Figure Figure 4

• 0 18r-------------------------------------------------------~

"in 16 18 16 16 -

14 U- -

ToO.ooso*

To .OO50-Range Of Dimensional 12 FMa Whew Leaage Has 12 4ý 11 11 Been Detected 10 10' 10 10" 10 10 t 899 8

~':" ~:~

C. a 14 a. 7 U. llean At Mean -o.oooa.

FiU -40005" 138 Measurements 138 Meaeurement8 6 6 I! i1I,:I 6~

6 4-2+

2 0.l1, I 011:1-Ii 0 ID fl 4 - . -

Dimensional Fit (inches)

Distribution of Dimensional Fits in DBNPS RPV Head NRC036-03676 NRC036-03676

Docket Number 50-346 License Number NPF-3 NPF-3 Serial Number 2731 2731 Attachment 22 Page 1 of 1I Key Information

• M EPRI EPRI Susceptibility Susceptibility Determination Criteria:

Davis-Besse Nuclear Power Station

.14.7 Lifetime EFPY '14.7 J RPV Head Temperature 605.00o F 605.0 I

• a CRDM Nozzle Information I

Number of Nozzles 69 Inside Diameter Diameter 2.765 in.

Outside DiameterDiameter 4.001 in.

Minimum Distance Between Nozzle Centerlines I 12.15 in.

I Arrangement - 69 CRDM Nozzles Nozzle Arrangement

• LIFTING Lues 13 LIFTINC STUD HOLE NUMBERS (1 THRU 60)

II'

[.114 *

~

/ 55

~9

15 n

56 lOT 125 45 21 JT IT 28

~,

50 5' 35 IS ., 10 ~

41 19 5

• 2 t4 31 GUIDE STUD

.... ~-

42 2~

I,.... 1- -f*

.. 3

. 6-15

~~ .

3t GUIDE STUD GUIDE STUD 3<4  !,

LOCATIONS LOCATlONS NOS.

NOS. 1515 88 ~5 e

45 " s.. 'Ie 12 IT'

" 11

.1 33 i23

" 32 27 4'

51 51 ., I 40 52

"" 55 '59

'" ~/

8 .. -I I/c"DlA.

84-1/ "DIA.

1A.BOLT B.C.HOLES BOLT HOLES ON 144 2.0 DIA.

ON ,",4'1z" B.C.

INOT INOT PRESSURE PRESSURE RETAINING RETAINING BOLTING)

BOL TINC) I Y

REACTOR REACTOR VESSEL CLOSURE CLOSURE HEAD KEY PLAN BOLTING DETAILS FIGURE rlGUR~

NRC036-03677 NRC036-03677

Docket Number 50-346 License Number License NPF-3 Number NPF-3 Serial Number 2731 Attachment Page 11 of2 of 2 0

COMMITMENT LIST The following list identifies those actions committed to by the Davis-Besse Nuclear Nuclear Power Station (DBNPS) in this document. Any other other actions discussed discussed in the submittal represent represent intended intended or planned planned actions by the DBNPS. They are described only for information and are not regulatory commitments. commitments. Please notify the Manager Manager -- Regulatory Regulatory Affairs ((419)-321-8450)) at the DBNPS of any questions regarding this document Affairs <<419)-321-8450)) document or associated regulatory commitments.

COMMITMENTS COMMITMENTS DUE DATE For the 13RFO, 13RFO, a qualified visual examination examination will be 13 th RFO 13" performed. performing this task will be instructed on perfonned. Personnel perfonning the type of unacceptable unacceptable conditions conditions using ONS3 as the basis.

Inspections Inspections will will be performed performed in in accordance accordance with with a procedure procedure developed developed specifically specifically for these examinations examinations that will meet the requirements of an ASME VT-2 basic requirements inspection. The previous VT-2 inspection.

inspection video of the cleaned head will be used to help determine unacceptable conditions. The RPV head will be detennine any unacceptable cleaned (as cleaned (as necessary) necessary) and and videotaped videotaped prior to to return return to to service service to re-establish to re-establish aa baseline baseline for for future future inspections.

inspections.

The acceptance acceptance criteria to be used for the qualified qualified visual inspection 13m RFO will consist of will of comparative comparative evaluations evaluations ofof any any as-found boric acidacid crystal deposits to photographs crystal photographs of leaking CRDM CRDM nozzles nozzles observed observed at ONS3 and Arkansas ONS3 Arkansas Nuclear One-Unit (ANO-1) and evaluation One-Unit 1 (AND-I) against any identified leaking CRDM nozzle flanges.

The DBNPS plans to perform inspections inspections of the RPV head and CRDM 13u, RFO nozzles as recommended recommended by MRP-48. The inspections will consist of MRP-48.

qualified visual inspections of the top RPV head bare bare metal surface at 13'hth RFO scheduled for the spring of 2002. If any leaks are the 13 detected, the source will be determined, the cracks leading to the leak leak characterized characterized by supplemental supplemental examination and the nozzle will be repaired.

The FirstEnergy FirstEnergy Nuclear Nuclear Operating Operating Company (FENOC)(FENOC) proposes to January anuary 29, 2002 2002 provide a final fmal response response to NRC Bulletin Request 3.a by January January 29, 2002 (60 days before 2002 before the start start of of 13RFO 13RFO scheduled scheduled forfor the the spring spring of of 2002). Final plans will 2(02). will be based on the be based the inspection inspection results results from from other facilities, the facilities, the ongoing ongoing work work ofof the the MRP, andand the the advancement advancement of Non- .

Destructive Examination Examination (NDE)

(NDE) technology technology and development development of of remote tooling remote tooling adequate adequate to to perform effective and perform effective and timely timely surface or surface or volumetric examinations from volumetric examinations from underneath underneath thethe RVP head.

RVP head.

NRC036-03678 N RC036-03678

Docket Number 50-346 License License Number NPF-3 Serial Serial Number Number 2731 e.

Attachment Attachment 3 Page 2 of 2 Page2of2 COMMITMENT LIST (continued)

COMMITMENT COMMITMENTS COMMITMENTS DUE DATE Details of the qualified qualified inspection plan will be developed prior to the March March 30.30, 2002 13RFO.

13RFO.

The DBNPS will continue to perform qualified visual examinations of perfonn qualified Ongoing the RPV head for evidence of leaking leaking CRDM nozzlesnozzles at subsequent subsequent refueling outages. The visual examination examination procedure procedure will be updated, as required,

. to include ,'industry f!Irlnl~h'v experience.

Flaws identified identified by NDE methods during the CRDM nozzles Ongoing requirements will be evaluated in inspections that are beyond current ~uirements accordance with the flaw evaluation accordance evaluation rules for piping contained contained in Section XI of the ASME Code The DBNPS will provide the NRC with the following information information 30 days following end of of within 30 days after plant restart following the l31b 13'h RFO scheduled scheduled to 13"111 RFO 13 begin in the spring of 2002:

a. A description description of the extent of RPV head nozzle leakage leakage and cracking, if detected.

cracking. detected. This information information will include the number, location, size and nature of each crack crack detected.

e b. A description description of the inspections requirements, requirements.

corrective corrective and actions inspections (type, scope, acceptance scope. qualification acceptance criteria). repairs and other taken criteria),

to satisfy applicable other applicable regulatory requirements.

requirements.

e NRC036-03679 NRC036-03679

1((lWin~ 11¥t-M-19/-a-~

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• NRC036-03 NRC 6 80 036-03680

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r-----.:~~~--=:-=-:_:_::------oo--------........::...=-J In re DAVID GEISEN ...n ~ fT exhibit ' ....- = . _

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SUMMARY

(Log No..# Sb, 1A-05-052 IR~~~~~~~~~~~~"~C:!!ir~cu~m!!!f~e~re!!:!n~tia~I..9!:~~~O~f

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Response to NRC Bulletin 2001-01, "Circumferential Cracking of ~R~ea~( Rea( Docket # 1A-OS-OS2 I~/ 3 Date Marked for ID~ 2008 (Tr. p.

,8 zS' p.- . .)

(4) commTMENT LUST ADDED TO LETTER 0 [(5 ) p Date Marked for lD:/ ,2008 IYESYES 1'1/9 116 76iiiiATERESPoN'SeDUETOBe~iMiTi:a;;=o;:;jf;;C------l7iSPi~

(6) DATE RESPONSE DUE TO BE SUBMITTED TO NRC 7_*EC Date Offered in Ev: ~I 11L __I,2008 (Tr. p.~-) p...2,&6_1.

Target.Date 8/30/01; Required 9/4/01

~~~~~~~~~~~~~--------~~~--~~

E]N/A rKEXP Through Witn . 9 anel:l Witne anel: MI ~ .IN fJ FI (8)PREPARED BY 1(9)NOTA1 Rod Cook ext. 7782 :YES Action: ADMI1TTý ITT REJECTED WITHDRAWN WITHDRAWN tt(1f111)~A~DO~m~o~NiAiAlLReFEfi:ENC:ES.......-

,11)

ADDITIONAL REFERENCES _ _ _ _..;;.;ext..;::...;77~8.;::2----_ _--.z.;;=-IiI. oatS: 2,,-~-,2~008 (Tr. p.

Date; 2/= ,2008 (Tr p.2 8V-t ) )

DOCKETED DOCKETED USNRC

~~~~M~M=~~S------~~~--------~~~~--~----~~------~~--~Sep.mb~~2009(11~Dam)

(13) COMMENTS September 9, 2009 (11:00)am)

Approvals continued on Page 2 OFFICE OF OFFICE OF SECRETARY RULEMAKINGS AND RULEMAKINGS ADJUDICATIONS ADJUDICATIONS STAFF STAFF (14)

(14) REVIEW AND

. REVIEW APPROVAL AND APPROVAL INITIALS DATE RECQVED APPROVED 0181 COGNIZANT REGULATORY REGULATORY AFFAIRS INDIVIDUAL INDNIDUAL R.M R.M...Cook *2,,oI 6/2,0(

0181 RESPONSIBLE RESPONSIBLE ENGINEER ENGINEER - MECHAN/CALDE.SIGN MECHANICAL DESIGN *P. Goyal P. Goyal ' --"g/Qi 2/-/L. */z !oi 0 RESPONSIBLE 181 RESPONSIBLE SUPERVISOR SUPERVISOR - MECHJSTR DESIGN MECHlSTR T. Swim Swim .n 0 RESPONSIBLE 181 RESPONSIBLE ENGINEER ENGINEER - PLANT PlANT ENGINEERING A.J. :Slernaszko A.J.Slemaszko £( ~fb..L..t 0 RESPONSIBLE 181 RESPONSIBLE SUPERVISOR-SUPERVISOR - PlANT PLANT ENGRG MECH ENGRG MECH J.B. cUMings Cunnings .('

~.~~

-LL,. I 181S.RESPONSIBLE RESPONSIBLE WORK WORK CONTROL INDIVIDUAL CONTROL.INDIVIDUAL M. McLaughlin Sjl'lESPONSIBC2 "BRIE PRBe et:l~bT

]*l~rLA'r~l*L i Zll-Lltr~ R~ ...... -ih*, r~v *='='i Ia1i SUPERVISOR, SUPERVISOR, DB DS UCENSING UCENSING Wuokko D.R. WUokko . ,/2 o.

1K SUPERVISOR.

KI SUPERVISOR, DB DB COMPUANCE COMPLIANCE D.L D.L Miller

• 40)-'(

0181 MANAGER, MANAGER, REGULATORY REGULATORY AFFAIRS AFFAIRS D. H. Lockwood D.H. B-, -I / */-  !

0VICE.PRESIDENTr l DATE ADDED TO LETTER (17) ADDITIONAL DISTRIBUTION DATE SENTTO NRC DATE SENT TO NRC

, 0, I, M " *)e*.vn

. DATE DATE OF OF BLIND BUND DISTRIB k*C ,)

6 £\n M. Pm r .

I I 511-00592 Sl1-00592 Exhibit Exhibit 87

\::S ~ Page 1I of Page of 12 NRC027 NRC027-1692 -1692

L..ETTERS - REVIEW NRC LETTERS REVIEW AND APPROVAL APPROVAL REPORT REPORT ED 7159-7 (1) RECORDS (1) MANAGEMENT NO.

RECORDS MANAGEMENT NO. 1~2)

T(2) SERIAL

- .- . 12731 SERIAL NO.

Paue 2 2731 Page2 (3)

SUMMARY

(Log No., Tte (3):

SUMMARY

Subject).

TItIeSybject)

Response to NRC Bulletin 2001-01.

2001-01, "Circumferential

• Circumferential Cracking of Reactor Reactor Pressure Pressure Vessel Head Head Penetration Penetration Nozzles-Nozzles" (4)COMMrrMENT LIST ADDED TO LETTER (5) PERIODIC II NON.pERIODIC NON-PERIODIC REPORT REPORT (4)COMMm.tENT lIST ADDEO TO l.ET'TER t8I (5) PERIODIC DYES 0iYES 0 t81NO NO REPORT NO._ NO. _ _

(6) DATE RESPONSE (6).DATE RESPONSE DUE Target Date 8130/01; DUE TO BE SUBMITTED SUBMITTED TO NRC Required 9/4/01 8/30/01; Required 9/4101 0N/A ON/A 7)

7) SPECIAL D

E EXPRESS HANDLUNG SPECtAL HANOUNG DELIVERY 0 EXPRESS DELIVERY t8I TELECOPY TELECOPY I DATE SENT DATE SENT (8) PREPARED BY (8) PREPARED BY (9) NOTARY (10)LICENSE FEE 1(10)LlCENSE REQUIRED FEE REQUIRED Rod Cook Rod extn82 ext 7782 0 YES [ NO t81vesDNo 10 DYES YES ~NO ONO (11) ADDITIONAL REFERENCES (11)ADDmoNAL REFERENCES (12) COMMITMENT NO.(S) CLOSED (12) COMMITMENT (13) COMMENTS (13) COMMENTS See Page Page 1 (14) REVIEW REVIEW AND APPROVAL AND APPROVAL INITIALS INITIALS DATE DATE RýCEIVED RECEIVED APPRO~EQ.

APPRLOVED 18I DESIGN ENGINEERING 0 DESIGN ENGINEERING MANAGER MANAGER D.C. Geisen Gelsen 181 0 PLANT ENGINEERING MANAGER MANAGER D.L Eshelman D.L Eshelman

~ A 0 WORK .CONTROL 18I CONTROL MANAGER MANAGER C.D.Nelson C.D. Nelson-" -:(7 fit. C.{J,A.J~ IJ -2..~V'1 3-z~"O')

.....~

0 DIRECTOR, 18I DIRECTOR. Wo.RK WORK MANAGEMENT MANAGEMENT J. Messina Messlna

~ -;or -01 't -""1 .-cJ I 18I 0 DIRECTOR, DIRECTOR, TECHNICAL TECHNICAL SERVICES SERVICES S.P.Moffitt S.P. Moffitt 0 DIRECTOR, 181 DIRECTOR, NUCLEAR SERVICES SERVICES LW.Wodey LW. Wodey 0

0 .

13 0

0]

0 ,

13 0

0i 0]

0 01 0I DATE DATE ADDED ADDED TO LETTEr TO LErTr~ (15) DATE ADDED ADDED By BY (17) ADDmONAL ADDITIONAL DISTRIBUTION DISTRIBlITlON DATESENTTONRC ~()

DATESENTTONRC 0 )f ~~Dl1J l DriDS; BY

~

BUNDDISTRIB~

DATE OF BLIND DISTRIBUT1 (.A,_

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\ BY k SI 511-0 0594 I .00594 NRC027 -1694 NRC027-1694

NRC L~ERS LET*ERS *- REVIEW REVIEW AND APPROVAL APPROVAL REPORT 1:07159-7 ED 7159-7 .

(1) RECORDS MANAGEMENT NO. NO. (2)SERIAL NO.

(3)

SUMMARY

(Log No., Tide Subject).

12731 Paae 1 1 I Response to NRC Bulletin 2001-01, "Circumferential Cracking. ofReactor Pressure Vessel Head Penetration Nozzles_

(4) COMMITMENT LIST ADDED TO LETTER

[I YES 0ONO (5) PERIODIC / NON-PERIODIC REPORT REPORTNO.

(6) DATE RESPONSE DUE TO BE SUBMITTED TO NRC 7) SPECIAL HANDLING DATE SENT Target Date 8/30/01; Required 9/4/01 El N/A 0 EXPRESS DEUVERY 0 TELECOPY (8) PREPARED BY (9).NOTARY (10) LICENSE FEE REQUIRED Rod Cook ext. nS27782 10 YES [I NO D0 YES 0 NO (I1) ADDITIONA' Oo cu... (12) COMMITMENT NO.(S) CLOSED

_______~ 6>en ~ V eIII___

(13) COMMENTI (14) INITIALS INmALS DATE

' RECEIVED APPROVED 181 COGN_ " **** __ v........"......

COGN......... Vt I"' ,.." I I"'\.M~

^ It~U.l1 Y IUUI'\L.

g,4usIuumI m.m.

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181 0 RESPONSIBLE RESPONSIBLE ENGINEER ENGINEER - MECHANICAL MECHANICAL DESIGN P.

P. Goyal 181 0 RESPONSIBLE RESPONSIBLE SUPERVISOR SUPERVISOR - MECHlSm DESIGN MEG-V/STR T. Swim 181 0 RESPONSIBLE RESPONSIBLE ENGINEER ENGINEER - PLANT ENGINeeRING ENGINEERING A.J. Sienaszko A.J.Sleinaszko 181 RESPONSIBLESUPERVISO~

0 RESPONSIBLE SUPERVISOR - PLANT ENGRGENGRG.MECH MECH J.B. Ci.Jnnlngs Cunnings 1810RESPONSIBLE RESPONSIBLE WORK CONTROL CONTROL INDIVIDUAL INDIVDUAL. M. Mclaughlin M. Mclaughin _______* ~

0 RESPONSIBLE (8J RESPONSIBLE WORK WORK PROD PROD StJPDT SUPDT D.E.Misslg D.E. Misslg 0 RESPONSIBLE LICENSING (8J.RESPONSIBLE UCENSING .INDIVIDUAL INDIVIDUAL F.W. Kennedy F.W.

o0 SUPERVISOR, SUPERVISOR, DB UCENSING UCENSING D.i=t D.R. Wuokko o0 SUPERVISOR, SUPERVISOR, DB DB COMPLIANCE COMPLIANCE D.L .ELMlaer Miller 181 0 MANAGER, MANAGER, REGULATORY REGULATORY AFFAIRS AFFAIRS D. H.H. Lockwood Lockwood 0

181 VICE PRESIDENT PRESIDENT CampbelU G.G. campbell

. . . . . . _. _ - r . . . . .. . .

• L. .. .

(15) DATE ADDED BY (17) ADDITIONAl DISTRIBUTION ADDITIONAL DISTRIBUTION DATE ADDED TO LETTER [I* I EATE SENT TO NRC " \p ( c .

DATE OF BLIND DISTRIBLJTI*ý*&ý M S()

Sl1-00595 511-00595 NRC027-1695 NRC027 -1695

NRC LF.TTERS LFTTERS -* REVIEW REVIEW AND APPROVAL APPROVAL REPORT t::0715~7 ED 7159-7 (1) RECORDS MANAGEMENT MANAGEMENT NO. '(2) SERIAL NO.

12731 Paae 2 (3) ~U"'IMAI~T No.. Title

SUMMARY

(Log No., TItle Sublect)

Subject) ' "

Response to NRC Bulletin 2001-01, uCircumferential Cracking .of Reactor Pressure Vessel Head Penetration No, 7lp.*

RA~~nnln~A to NRC Bulletin 2001-01 (4) COMMITMENT LIST ADDED TO LETTER [ (5) PERIODIC / NON-PERIODIC REPORT 0 YES 0 NO REPORT NO.

(6) DATE RESPONSE DUE TO BE SUBMITTED TO NRC 7) SPECIAL HANDUNG DATE SENT SENT Target Date 8/30/01; Required 914101 El N/A 0 EXPRESS DELIVERY 0 TELECOPY PREPARED BY (8) PRF"PAIRED (9) NOTARY (10) LICENSE FEE REQUIRED Rod Cook ext 7782 0YES [I NO [] YES ONO (11) ADDITIONAL REFERENCES (12) COMMITMENT NO.(S) CLOSED (13) COMMENTS See Page 1 (14) REVIEW AND APPROVAL REVIEW AND APPROVAL INITIALS DATE

____RECEIVED, APPROVED 0 DESIGN ENGINEERING I8IJ;lESIGN ENGINEERING MANAGER MANAGER D.C. Geisen //2/

181 0 PlANT PLANT ENGINEERING ENGINEERING MANAGER MANAGER D.L. Eshelman Eshelman 181 0 WORK CONTROL MANAGER MANAGER C.D. Nelson C.D. Nelson 181 DIRECTOR, MANAGEMENT DIRECTOR, WORK MANAGEMENT J. Messina J. Messlna --# - -.

181 DIRECTOR, DIRECTOR, TECHNICAL TECHNICAL SERVICES SERVICES S.P.

S.P. Moffitt 4/

1810DIRECTOR, DIRECTOR, NUCLEAR SERVICES NUCLEAR SERVICES L.W. Worley L.W. Wolley goe o

o o0 o

o0I 01 DATE ADDED TO LETT'ER [3 (15) DATE ADDED BY (17) ADDITIONAL DISTRIBUTION DATE SENT TO NRC ()TB DATE OF BLIND DISTRI N TRIB S11-00596 511*00596 NRC027-1696 NRC027 -1696

The NRC Letters - Review and Approval Approval Report Report (ED 7159-7) should should be completed by the Regulatory Regulatory Affairs Section.

BLOCK 1 MANAGEMENT NO. - Regulatory Affairs enters Records RECORDS MANAGEMENT Records Management Management number number prior to correspondence to NRC.

distribution of correspondence.to BLOCK BLOCK 2 SERIAL NO.NO. - Initiator enters serial number number obtained from the Regulatory Regulatory Affairs Affairs Clerk.

BLOCK 3

SUMMARY

(Log No., Title Subject) - Initiator enters a summary of the correspondence.

SUMMARY

correspondence. This summary should identify ififthe correspondence correspondence is in response to any previous previous correspondence correspondence and why the letter is being written.

BLOCK 4 COMMITMENT LIST ADDED TO LETTER - Preparers checks the block COMMITMENT block to indiate a commitment commitment list has included with the letter.

been included BLOCK 5 PERIODIC/NON-PERIODIC REPORT PERIODICINON-PERIODIC REPORT - Identity Identify whether this correspondence correspondence is a Periodic Periodic or Non-Periodic Report Report as identified in Nuclear Group Procedure Procedure NG-NS-00807.

NG-NS-00807.

BLOCK 6 BLOCKS RESPONSE DUE TO BE SUBMITTED DATE RESPONSE SUBMITTED TO NRC - Initiator enters the date the correspondence correspondence is is due to the NRC. If If the correspondence correspondence does daes natnot have a required due date, the block shall be marked marked not applicable applicable (NA).

BLOCK 7 SPECIAL HANDLING HANDLlNG-- Initiator checks if if the correspondence requires the.torrespondence requires special special distribution to the NRC. IfIf yes, the Regulatory Regulatory Affairs clerk enters date the correspondence correspondence is sent.

BLOCK 8 PREPARED BY - Initiator enters the names of individuals responsible for providing technical information BLOCK 8 o PREPARED BY -Initiator enters the names of individuals responsible for providing technical information for the correspondence correspondence alang along with his/her name.

BLOCK 9 NOTARY - Initiatar NOTARY Initiator checks ififthe carrespandenceis correspondence is required to be natarized. notarized.

BLOCK 10 LICENSE FEE REQUIRED REQUIRED - Initiator checks ifif a license license fee is required, per the requirements of af 1iG Part 170.

170. IfIfyes, the initiator shall campletecomplete a Voucher Voucher Check Authorization Autharization (Form 294) and obtain*

appropriate fees to.

appropriate to accampany accompany the correspondence.

thecarrespondence.

. ~'.

BLOCK 11 ADDITIONAL REFERENCES ADDITIONAL REFERENCES - Initiator enters any additional NRC correspondence correspondence or ar documents that to the subject correspondence.

pertain to. correspondence.

BLOCK 12 COMMITMENT NO(S).

COMMITMENT NO(S). CLOSED - Initiatar Initiator enters the Commitment Management System number(s) of Commitment Management commitments that are closed any. cO(11mitments clasedby by the subject subject correspondence.

correspandence.

. ~. .

BLOCK 13 COMMENTS -Jnitiatot

. COMMENTS -Jnitiator or any reviewer enters appropriate approprriate comments regarding regarding the subject correspondence.

carrespondence.

BLOCK 14 REVIEW AND APPROVAL APPROVAL - Initiator Initiator checks and and/or enters the desired

/arenters desired reviewer(s).

reviewer(s). The technical accuracy of a respanse accuracy response to the NRC NRC is the responsibility respansibility of the Director and Management Management individual assigned the action.

actian.

BLOCK 15 DATE ADDED BY - Distributar Distributor checks the Date Date Added to.to Letter Block Block and signs the Date Added By By block to indicate indicate the original letter been dated prior to distributian letter has be~n distribution to the NRC.NRC.

BLOCK 16 DISTRIBUTED DISTRIBUTED BY - DistributionDistribution to. to the NRC NRC shall be made by the Regulatary Regulatory Affairs Section. Distributor Affairs Sectian. Distributar signs the Distributed By block block and completes campletes the Date Sent to. to NRC block.

BLOCK BLOCK 17 17 ADDITIONAL ADDITIONAL DISTRIBUTION - Initiator ~nters enters individuals individuals requiring distributian distribution that are not on the the

.standard standard distribution distribution list.list.

BLOCK 18 DISTRIBUTED DISTRIBUTED BY - Distributor signs the Distributed By block arnd comp.letes the blocK andcampletes .the Date of Blind Distribution block.

Distributian  : , ; _00 -- - .... : :

I 0

\ '!

Si1-00597 511*00597 NRC027-1697 NRC027-1697

. \

RJT?-C(~

(

DOCKETED USNRC USNRC FirstEnergy FlrstEilerfly . ~

September 9, 2009 (11:00am)

September (11 :OOam)

Dais-Besse Nudear DBv;s.,Besse Oak Harbor Oak PowerSlaliofl Nuclear Power 5501 Norfh State 550J NoM*

Hamor. Ohio Station Route 2 STare Route Ohio 43449."760 43449-9160 2

SECRETARY OFFICE OF SECRETARY 419-321-588 RULEMAKINGS AND RULEMAKINGS AND - 4f~32'-6588

':>uy 3uyG. G *.CampbeII CBmptJeJl Vice President Nuclear President - Nuclear ADJUDICATIONS STAFF ADJUDICATIONS I fJ ~~ th Fax: 419-321-8337

,23

":S Docket Number 50-346 Docket 50-346 U.S. NRC U.S. NRC.

In rfa DAVID

. In GEISEN ~"- r-r DAVID GEISEN 5

c. i r-r Eit 0 I{

ExtIIbIt .-:..:.--

Number NPF-3 License Number Docket ##1i1A-05-052 Docket A-05-052 Ii'} Jy ~ [5 Serial Number Markedlor lD~

Date Marked10r ID 2008 (Tr. P p. . .),)

Serial Number 2735 October 17.

October 17, 2001 Date J#.L.

Ofterecl in Ev: ___ 2008 Date Offered p. ~& )

(Tr. p.

2008 (1'r. g g Z )

Through hough Witness/Panel:

Witnes/P ne:N/ pc /, .

u.s. Nuclear Regulatory Commission U.S. ActiOn:

Ac" ,~ REJECTED REJECMD Wn1TDWAWN WITHOAAWN Document Control Desk Attention: Document ~~(I'r.p.S2.--0 )

Washington, DC. 20555-0001 20555-0001 Subject; Supplemental

Subject:

Supplemental Infonnation Response to Information in Response to NRC Bulletin 2001-01, 2001-01, Cracking of Reactor Pressure Vessel Head "Circumferential Cracking "Circumferential Head Penetration Penetration Nozzles" .

Ladies and Gentlemen:

attached provides The attached provides supplemental information concerning the Davis-Besse Nuclear supplemental informationconceming Nuclear Power Station, Unit 1 (DBNPS) ~ponse (Senat (DBNPS) response (Seial Number 2731, September 4, 2731, dated September Regulatory Commission (NRC) Bulletin 200.1-01, 2001) to Nuclear Regulatory 2001) "Circumferential 2001-01. "Circumferential Reactor Pressure Cracking of Reactor Pressure Vessel Head Penetration Nozzles."Nozzles." Portions of this information were discussed with members members of the NRC staff on October October 3 and 11, 11,2001.

.200 1.

DBNPS and NRC staffs are scheduled In addition, the DBNPS scheduled to to meet and discuss this information and additional NRC information NRC crack growth information on October 24, growth modeling infonnation 2001.

2001. .

This submittal submittal provides updated and additional provides updated additional information in support oithe of the basis for Davis-Besse Nuclear Power Station (DBNPS) until operation of the Davis-Besse the continued safe operation its next scheduled refueling outage commencing in outage commencing in March 2002, at which time the Control Rod Drive Mechanism (CRDM)

Mechanism (CRDM) nozzles Reactor Pressure Vessel (RPV) and Reactor penetrations will head penetrations inspections or appropriate Will undergo qualified visual inspections supplemental appropriate supplemental inspections.

inspections.

refueling outage, the RPV head was inspected..

1996, during a refueling In May 1996. inspected. No leakage was identified, and these results have been recently verified by aa re-review identified. re-review of the video video tapes obtained from that inspection. The RPVheadRPV head was mechanically cleaned at the mechanically cleaned end of the outage. Subsequent inspections of the RPV bead head in the next two refueling refueling (1998 outages (1998 and 2000), also did not ~91eak:age leakage in the CRDM nozzle-to-head CRDM nozzle-to-head areas that could be inspected. Video tap~~~~tapefsen during these inspections have also inspections have also been been re-reviewed. September 9, 2009 (11:00am)

September 9, 2009 (11 :OOam)

OFFICE OF SECRETARY RULEMAKINGS AND RULEMAK1NGS ADJUDICATIONS STAFF ADJUDICATIONS STAFF Exhibit 198 Exhibit 198 Page 1 of 111 Pages NRC004-1156

'N'T, ý(_- slý-" .0-'Vý \)S~ 111 Pages NRC004-1156

Docket Number 50-346 50-346 License Number NPF-3 NPF-3 Serial Number 2735 2735 Page 2 of2 of 2 Accordingly, using the end of the outage outage in 1996 as in 1996 as the postulated worst-case time for an axial crack to reach a through-wall condition, the projected time for the crack to ~achits reach its critical circumferential size;~i1$

through-wall circumferential sizewas..determined

~termined based on the results from an FramatlJIIleF-ramatome ANPANP

,, 1a~tsment- This RWHead:

i:.Il_~ment RWV!Read NQzzleand Nozzle and Weld Safety Assessment demonstrates demonstrates the postulated crack will take apprc:iXil'nately:7,5:years'to crack approxmately7.5:,years'to manifest .into into an ASME Code allowabJe allowable crack size.

SA~ppying this 7.5 years.to the May_1996 inspection projects the worst-case allowable crack size

, A.p,pJY,ipg this 7.5 y~Jo .th~,~ay)~96 inspection projects the worst-case allowable crack size

.beingreiched in November

.... "oeingreached Novembýeira2003. 2 ý ItI't'is important to note the is imponant the allowable crack size will still rn.. .*ntain.

an ASME

.. _: ... ;maintain:ap.ASMECode safeto/Code safety factor of three.

~; Jr(" ; ~'~ ... ' ,~.~.;./...~' ~\ .* t.:.~,:(~-

rift Element Gap

.. --'~A'FiriiteElement

-'-- Gthp Analysis was peIformed peiformed by Structural Integrity Associates to verify the

"~~J1?gapsli~'rtVec!n;the

• p'bg ~hse CRJijjM~ilb:iZles:8htFthe CRDM~tizles ahdMthe RVP head during nonnal normal operation oDeration would pennit permit thr~l:lgh:}'f.a.J1.1~f1g~Jr.tW~~i.~Q~r througfh-all laagq,,frm aynozT. r through-weld through-weld cracks cracks in the the J-groove weld to J-groove weld to be be observed via boric acidcrystaI acid crystal deposits. This analysis concluded concluded that all but four nozzle/penetration interfaces nozzle/penetration interfaces would show visible leakage. These four nozzles are in the least stressed stressed. area of the RPV head, and where no leakage attributed to drcumferentialcracks circumferential cracks has been observed at any other been observed ~ any other plants. plants.

The DBNPS staff is continuing to be involved in and monitoring industry developments developments regarding CRDM nozzle/penetration nozzle/penetration cracking, modifying its inspection cracking, and .modifying inspection plans as appropriate.

previous inspections conducted.

Based on the previoUs conducted, re-~viewed re-reviewed inspection videos.

videos, analyses that have been perfonned performed concerning concerning crack growth rates. rates, the the ability to identify cracking. and industry cracking, evaluations evaluations and findings, it is concluded concluded there is reasonable assurance that theDBNPS the DBNPS will continue to operate safely to the next refueling outage scheduled for March 2002.

outage scheduled If you have any question or comments, please contact Mr. David H. Lockwood, Manager, H.Lockwood, Manager.

Regulatory Affairs, Regulatory Mfairs. at at (419)

(419) 321-8450. 321-8450.

Very truly yours, Enclosure Enclosure and Attachments cc: J. E. Dyer, Regional Regional Administrator, Administrator, NRC NRC Region RegionID M11 S.

S. P. Sands, Sands. DB-l DB-l NR~ NRC_3R Project Manager Manager D. Simpkins, DB-l DB-l Acftg~g Senior Resident Inspector Inspector L.A.

J.A. Zwolinski, Zwolinski. NRC/NR1Uk NR~Director. Director, Licensing Project Management Project Management S.S. Bajwa, S.S. NRCINRR Director, Project Directorate m Bajwa, NRC/NRR HI AJ.

A.J. Mendiola, Mendiola, NRC/NRR NRC/NRR Chief, Chief. Projects Projects Section 11-2 Sectionm-2 Utility Radiological Radiological Safety Board NRC004-1157 NRC004-1157

i RASX ~/~'

\ '

_. '\

I 41r-ýVly IT FirstEnerg FlrstEherSY , ~

.pe Davis-Besse Nudear Da\lis-,Besse 5501 5501 NOf1fI Oak Hartor.

PowerStaion Nudear Power No'U St8te HadXlr. O/Iio S/allon State.Route

,Route 2 Otho 43449-9760 43449-9760 2

t3uy G.campbell G. campbell , F 419-321-=58 419-32HJ588 President-Vice Ptesident - Nuclear Nuclear I tv n~ rt-, Fax: 41rJ.,32t-8337 IV>

~

Docket Number 50-346 50-346

~!

Inre ~:D U.S. NRC GEISEN GEI DAVIDGESEN 5~v.. F+

5u ExhIbIt *#...;../~(

( _

Number NPF-3 License Number IA-05-052 Docket ## 1A'()S'()52 25 Serial Number 2735 Date Marked-for Date Marked10rlD*I"Z.)g 2008 (Tr. p.. .9 ID:-2-11-. ,2008(Tr. p.L_)

2008 (Tr. p.X 2- -Lo

-)

Date Otlered in Ev: 'ý'

Date Offered'in Ev: J1.jL. 2008 err. p.~ :?- 6; )

October 17. 17, 2001

'Through T Witness/Panel:

WREJETPa.

AMouTh N,I p.AJ P ,'

Regulatory Commission Nuclear Regulatory U.S. Nuclear ADMITTED REJECTED AdiOn: ADMITTD Acdom REJECTE WITHDRAWN WITHDRAWN Attention: Document Washington, Washington, DC.

Control Desk DocumentCoiltro}

20555-0001 00 0.: 1141-.2008

) 1-2 =00o~r.32S-b (fr. p.E;L & )

Information in Response to NRC Bulletin Supplemental Infonnation

Subject:

Supplemental 2001-01,Bulletin 2001-01.

Cracking of Reactor Pressure Vessel Head Penetration "Circumferential Cracking "Circumferential Nozzles" Ladies and Gentlemen:

attached provides supplemental The attached The concerning the Davis-Besse Nuclear supplemental information concerning 1 (DBNPS)

Station. Unit (DBNPS)

Power Station, response (Serial 2731, dated September 4, "

(Senat Number 2731, 2001) to Nuclear Regulatory Commission 20(1) 2001-01, "Circumferential Commission (NRC)Bulletin 2001-01. "Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles."... Portions of this Penetration Nozzles this information were discussed with members of the NRC staff on October 33 and 11.

information 11,2002001.1.

In addition, the DBNPS and NRC staffs are scheduled to meet and discuss this the.DBNPS this information and additional NRC crack growth modeling information on October information October 24. 24, 2001.

2001.

This submittal provides updated and additional information This information in support of in suppott ofthe the basis basis forfor Davis-Besse Nuclear Power Station continued safe operation of the Davis~Besse the continued (DBNPS) until Station (DBNPS) until its next scheduled refueling outage commencing in March 2002, 2002. atat which time time thethe Control Rod Drive Mechanism (CRDM) nozzles and Reactor Pressure Vessel (RPV)

(RPV) head penetrations will undergo qualified visual inspectionsinspections or appropriate supplemental supplemental inspections.

In May 1996. outage, the 1996. during a refueling outage. inspected. No leakage the RPV head was inspected. leakage was identified, and these results have identified. been recently bavebeen verified by recently verified a re-review of by are-review of the video the video obtained from that inspection. The RPV head was mechanically tapes obtained mechanically cleaned cleaned at the the inspections of Subsequent inspections end of the outage. Subsequent RPV head of the RPV bead inin the the next two refueling next two refueling 20(0). also did not identify any leakage in the CRDM nozzle-to-head (1998 and 2000),

outages (1998 areas that could be inspected. Video tapes taken during these inspections have also been re-reviewed.

been re-reviewed. ~

ry:'6 Exhibit 198 198

,fC. RL f4-.t(/c I- Page Page 1 of 111 111 Pages NRC004-1 NRC004-1156156

'~ .'-..

Docket Number Number 50-346 50-346 NPF-3 Number NPF-3 License Number Serial Number 2735 2735 Page 22of2of 2 Accordingly. using the end of the outage in 1996 as the postulated worst-case Accordingly, worst-case time for an axial crack to reach a through-wall condition. the projected time for the crack to reach through-wall condition, ~ach itsits critical through-wall circumferential si~~;~~~tennined through-wall circumferential sir;w .astennined based on the results from an Framatome Framatome ANP

..*~ lOS'essment-.-

l~S:e~~~~t._ This ~N~.ijead:No~eand~

RVHead N .zzle andtWeld Weld Safety Assessment demonstrates the postulated postulated crack will take approXiinatelym5;~ears:to apprdf m~telyt7v5*.years-to manifest into an ASME Code allowable crack size.

L _.

.Applying this

_~.\~1J?f!y.t.n~

~l:5ejng 6*jng reached al. reihed JD

?5

~is 7.5 years in Nov4nember to the May 1996 inspection

~1~,J9}!k~f¥.ay~J.~9~

November 2003. It .Islmportant inspection projects projects the the worst-case is important to note the allowable allowable crack worst-caseall~wabl~

allowable crack size wlUstill c~ksize will still size

_~~**Qma}ntainlan-AS~,eode*safet~dactor of three.

( .m...*aIntainflaý..AS$ME,.ode'safetiyfactor

(_,

f' ri"~~\)-('i.~;W fi"J~1"f:6[

-. ,.. -.==KFinite Eleindnt 'baR" Fihiti Element AiuilySis was performed Gap Analysis performed by Structural Integrity Associates to verify Structural Integrity verify the

~~mgilps(6aw~n;tlie Cgaps bti een-the CRDM_ CRE):t~i~fib'"iZles:ana-rthe R les andlthe RVP head during nonnal normal operation would pennit permit thro\~g~~!J.~~~g~1r9w~~Y"nQ.~l~;Qr through-wall leak:agr through-weld through-weld cracks in the I-grooveJ-groove weld to be observed via boric acid crystal deposits. This analysis concluded that all but four nozzle/penetration interfaces would show visible leakage.

nOzzle/penetrationinterfaces Jeakage. These four nozzles are in the least least stressed area of the RPV RPVhead,head, and where no leakage circumferential cracks has leakage attributed to Circumferential been observed observed at any other plants.

The DBNPS staff is continuing to be involved involved in and monitoring monitoring industry industry developments regarding CRDM nozzle/penetration nozzle/penetration cracking.

cracking, and modifying its inspection inspection plans as appropriate.

Based on the previous inspections conducted, re-reviewed inspections conducted. re-reviewed

-t, inspection videos, videos. analyses that have been performed concerning crack growth rates.

performed concerning rates, the ability to identify cracking, and industry identify cracking.

evaluations and findings, it is concluded there is reasonable reasonable assurance that the DBNPS will continue to operate operate safely to the next refueling outage scheduled scheduled for March March 2002.

If you have any question question or comments.

comments, pleaseplease contact Mr. David H.Lockwood, Manager, H. Lockwood, Manager.

Affairs, Regulatory Affairs.

Regulatory at (419) 321-8450.

Very truly yours.

Enclosure and Attachments Enclosure Attachments Administrator, NRC Regionm cc: J. E. Dyer, Regional Administrator. Region mI S. P. Sands. DB-l DB-1 NR~ NRCfW Project Manager Manager D. Simpkins, Simpkins, DB-l Actig Ac!p.g Senior Resident InspectorInspector J.A. Zwolinski. NRCINRRDirector, I.A. Zwolinski. NR~Director, Licensing Project Management Management S.S. Bajwa, NRCINRR S.S. Bajwa. NRC/NRR Director, Director. Project Directorate m III Mendiola, NRC/NRR A.J. Mendiola. NRCINRR Chief, Projects Sectionm-2 Section M-2 Radiological Safety Board Utility Radiological S

"* * '"-* *. A ).'~ NRC004-1157

.:: ~'l '?-'-4

• -'4* {. _-J ~;;

I /HI ',N.',RC, 004-1157

I Docket Number 50-346 Number 50-346 License License Number NPF-3 Number Serial Number 2735 Enclosure Enclosure Page I1 of 1I SUPPLEMENTAL INFORMATION SUPPLEMENTAL INFORMATION IN RESPONSE RESPONSE TO BULLETIN 2001-01 NRC BULLETIN FOR DAVIS-BESSE DAVIS-BESSE NUCLEAR POWER POWER STATION UNIT NUMBER NUMBER 1 I This letter is submitted pursuant 50.54(f) and contains supplemental pursuant to 10 CFR 50.S4(f) supplemental information information concerning concerning the response (Serial (Serial 2371, 2371. dated September September 4, 2001) 2001) to NRC Bulletin 2001-01,Bulletin 2001-01.

"Circumferential Cracking of Reactor Pressure Vessel Head Penetration Nozzles,"

"Circumferential Nozzles." for the Davis-Nuclear Power Station, Unit Number 1.

Besse Nuclear I, Guy G. Campbell.

1, Campbell, state that (I)

(1) I am Vice President - Nuclear of the FirstEnergy NuclearNuclear Operating Company, Company, (2) Ilamam duly authorized to execute execute and file this certification certification on behalf of Company and The Cleveland Efectric the Toledo Edison Company Ekctric Illuminating nJuminating Company, and (3) the statements set forth herein are true and correct to the best of my knowledge, statements information and knowledge. infonnation and belief.

Campbell, Vice President For: G. G. Campbell. President - Nuclear BT-~~

L."W. Worley, Directork-L.W. Worle~Services plx..Srt., ce Afftrm.ed and subscribed before me this 17th Affirmed 17th day of October, October, 2001

~~~&~

NtarYPUbliC.

Ntary Puc, State of Ohio Laura A. Jennison Jennison Commission Expires on August 16, My C01IIIIlission 16, 2006.

2006.

NRC004-1158 NRC004-1158

Docket Number 50-346 Number 50-346 License Number NPF-3 Number NPF-3 Serial Number 2735 2735 Attachment 1 I of 5 Page 10fS SUPPLEMENT SUPPLEMENTAL INFORMATION IN RESPONSE TO NRC BULLETIN AL INFORMATION BULLETIN 2001-01 The Davis~Besse Station, Unit II (DBNPS) submitted Davis-Besse Nuclear Power Station, submitted its response to NRC NRC 2001-01, "Circumferential Bulletin 2001-01, "Circumferential Cracking Pressur~ Vessel Head Cracking of Reactor Pressure Penetration Nozzles" Penetration FirstEnergy Nuclear Nozzles" in FlI'StEnergy Nuclear Operating Company (FENOC) letter Serial Serial Number 2371, dated September Number 2371, September 4, 2001. Portions of this information 4,2001. infonnation have have been discussed with members members of the NRC NRC.staff on October October 3 and II,11, 200 2001.1.

SUMMARY

This submittal provides updated and additional information in support of the basis for the additional infonnation operation of the Davis-Besse Nuclear continued safe operation continued Nuclear Power Station (DBNPS) until its next scheduled refueling outage commencing scheduled refueling commencing in March 2002. 2002, at which time the Control Mechanism (CRDM) nozzles and Reactor Rod Drive Mechanism Reactor Pressure Vessel (RPV)(RPV) head penetrations penetrations will undergo qualified visual inspections.

1996, during a refueling outage, the RPV head was inspected. No leakage was In May 1996.

identified. and these results have been recently verified by a re-review identified, re-review of the video tapes obtained from that inspection. The RPV bead

. obtained head was mechanically cleaned at the end of the mechanically cleaned the.

outage. Subsequent inspections inspections of the RPV head in the next two refueling outages oftheRPV (1998 outages (1998 and 2000), also did not identify any leakage in the CRDM nozzle-to-head nozzle-to-head areas that could could be inspected. Video tapes taken during these inspections inspections have also been been re-reviewed.

re-reviewed.

Accordingly, using the end of the outage in 1996 Accordingly, 1996 as the postulated worst-case time for an an axial crack to reach reach a through-wall projected time for the crack through-wall condition, the projected crackloto reach its critical through-wall circumferential size was determined through-wall circumferential determined based on the results from an Framatome ANP assessment. This RV .Head Framatome Head Nozzle and Weld Safety Assessment Weld.Safety AssesSment demonstrates the postulated crack will take demonstrates take approximately approximately 7.S7.5 years to manifest manifest into an allowable crack size. Applying.this ASME Code .allowable.crack Applying this 7.5 years to the May 1996 inspection inspection allowable crack size being reached in November projects the worst-case allowable projects November 2003. It is is important to note the allowable crackcrack size will still maintain an ASME anASME Code safety factor of three.

oftbree.

A Finite Element Gap Analysis was performed by Structural Integrity Associates to AFiniteEIement between the CRDM verify the gaps between CRDM nozzles and the RVP head during normal operation operation leakage from any nozzle or through-weld through-wall leakage would permit through-wall through-weld cracks in the I- 1-groove weld to be observed via boric acid crystal deposits. This analysis concluded concluded that all but fourD~eJpenetration interfaces would show visible leakage. These four nozzles four nozzle/penetration interfaces nozzles are in the least stressed area of the RPV head, leakage attributed bead, and where no leakage attributed to circumferential circumferential cracks has bas been observed atany at any other plants. .

The DBNPS continuing to be involved in staff is continuing DBNPSstaff and monitoring industry developments jnandmonjtoring nozzle/penetration cracking, regarding CRDM nozzle/penetration cracking, and inspection plans as modifying its inspection appropriate.

NRC004-1159 NRC004-1159

..4. .

Docket Number 50-346 License Number NPF-3 License Serial Number 2735 Attachment 1 Attachment Page 2 of 5 Page2of5 Based on the previous inspections conducted.

on.the conducted, re-reviewed re-reviewed inspection videos, analyses that have been performed concerning concerning crack crack growth rates, rates. the ability to identify cracking, and industry evaluations evaJuations and findings, fmdings. it is concluded there is reasonable reasonable assurance that the DBNPS will continue continue to operate safely to the next refueling outage scheduled scheduled for March 2002.

PLANT DESIGN The DBNPS has a BabcockBabcock & & Wilcox (B&W) nuclear steam supply system. The design is similar to other B& B&W 177-fuel assembly plants, except that DBNPS is of the raised-W I77-fuel loop design, design. The DBNPS DBNPShas has 69 Control Rod Drive .MechanismMechanism (CRDM)nozzles (CRDM) nozzles of which 61 are used for CRDMs, 7 are spare, and one is used for the Reactor Pressure Vessel Vessel (RPV) continuous head venL vent. Each CRDM nozzle is constructed of Inconel Inconel Alloy 600 and is attached to the RPV head by an Inconel Inconel Alloy 182 J-groove I-groove weld. The DBNPS is unique in the B&W fleet in that it is the only unit that.

fleetin that has aaRPV RPV head head continuous vent that allows for the movement movement of coolant around the interior interior of the head, head.

thereby minimizing minimizing the stagnation stagnation of hot coolant in the top of the head and trapping of air air or oxygen.

PREVIOUS lNSPECflON PREVIOUS INSPECTION RESULTS In FENOC FENOC letter Jetter Serial Number 2731, 2731. the past inspections of the DBNPS Reactor Reactor Vessel (RPV) head were discussed. As a result of NRC staff questions, Pressure Vessel supplemental supplemental information to and amplification amplification of that discussion discussion.isis provided in the following.

performed ouring The inspections performed during the lOlIl, ll th,and 1O'h, 11, and 12th (10RFO, 12 th Refueling Outage (lORFO, conducted April 8 to June 2.

conducted 2, 1996; 1996; liRFO, 1lRFO, conducted conducted April 10, to May 23, April10. 23, 1998; 1998; and, 12RFO, conducted 12RFO. conducted April 1I to May 18. 18, 2000) consisted of a whole head visual inspection ofthe RPV head in accordance accordance with the DBNPS Boric Boric Acid ControlProgram Control Program pursuant to Generic Letter 88-05.

88-05, "Boric Acid Corrosion of Carbon Carbon Steel Reactor Reactor Pressure Boundary Pressure Boundary Components in PWR Plants." Plants." The visual inspections were conducted by remote camera camera and included below insulation insulation inspections of the RPV bare head such that the Control Control Mechanism (CRDM) nozzle penetrations were viewed. During lORFO.

Rod Drive Mechanism 10RFO, 65 ofof 69 nozzles were viewed, vieWed, during 11RFO. 50 of 69 nozzles were viewed, and during 12RFO, 12RFO, 45 of69of 69 nozzles nozzles were viewed. It should be noted that 19 of the obscured obscured nozzles nozzles in 12RFO were also those obscured in in llRFO.Following I IRFO. Following IIRFO, llRFO. the RPV head was mechanically mecbanically cleaned cleaned in localized areas as limited by the service service structure structure design.

Following 12RFO, Following 12RFO. the RPV

.RPV head was cleaned cleaned with demineralized demineralized water to the extent extent possible to provide a clean head for evaluating future inspection possible to provide a c1eanhead for evaluating future inspection results. results.

The affected affected areas of accumulated accumulated boric acid crystal crystal deposits were video taped, and have subsequently been reviewed with specific focus on boric acid crystal deposits with subsequently N RC004-1160 NRC004-1160

Docket Number 50-346 Docket Number 5G-346 License Number NPF-3 License Number NPF-3 Serial Serial Number Number 2735 2735 Attachment Attachment 1I Page Page 33 ofof5 5 reference reference to to the the CRDM CRDM nozzle nozzle penetration leakage as penetration leakage as previously previously observed observed at at the the Oconee Oconee Nuclear Station, Unit NuclearStation, Unit 33(ONS-3)

(ONS-3) and and atat Arkansas NuclearOne, Arkansas Nuclear One, Unit (ANO-!). During Unit I1(ANO-1). During the 12RFO inspection, the 12RFO inspection, 24 ofthe 24 of the 6969 nozzles nozzles werewere obscured obscuredby boric acid by boric acid crystal crystal deposits deposits that were that were clearly clearly attributable attributable to to leaking motor tube leaking motor tube flanges flanges from from the the center CRDMs. A center CRDMs. A further subsequent review of the video tapes has been conducted and corroborates the further subsequent review of the video tapes has been conducted and corroborates the previous statements previous statements and conclusions stated and conclusions stated inin letter letter Serial Number 2731 Serial Number 2731 that that the the results results of this review did not identify any boric acid crystal deposits that would have been of this review did not identify any boric acid crystal deposits that would have been attributed attributed to to leakage leakage from from thethe CRDM CRDM nozzle penetrations, but nozz}epenetrations, w~re indicative but were indicative of of CRDM CRDM flange flange leakage.

leakage. IncludedlDc1uded as as Attachments Attachments 22 and and 33 are the inspection are the inspection resultsresults for 1ORFO, 10RFO, It..;..:: ~

IllRFO IRFO and and 12RFO, l2RFO. and and aa figure representing these figure representing these nozzle nozzle locations, locations. respectively.

~tively.

In summary. results In summary, from previous results from inspections of previous inspections ofthe the CRDM CRDM nozzle penetrations provide nozzle penetrations provide reasonable assurance for the continued safe operation reasonable assurance for the continued safe operation of the DBNPSuntilthe next of the DBNPS until the next refueling refueling outageoutage in in March March 2002.2002.

ANALYTICAL WORK ANALYTICAL WORK PERFORMED; PERFORMED~

RV Head Nozzle RV Head Nozzle and Weld Safet and Weld Safety Assessment Assessment Attachment 4, Attachment Framatome ANP's

4. Framatome ANP's non-proprietary document FRA-ANP non-proprietarydocumentFRA ..ANP 51-5012567-01, 51-5012567-01,

"'RV Head Nozzle uRV Head Nozzle and and WeldWeld Safety Assessment." provides Safety Assessment," provides an an assessment assessment that that demonstrates.safe demonstrates, safe operation operation of ofthethe Babcock&

Babcock"& Wilcox (B&W)-designed nuclear Wilcox (B&W)-designed nuclear steamsteam.

supply systems with supply systems with thethe potential potential for for primary primary waterwater stress stress corrosion corrosion cracking cracking (PWSCC)

(pwSCC) of RPV .head penetration nozzles. The document addresses the assumed presence of of RPV head penetration nozzles. The document addresses the assumed presence of PWSCC PWSCC in in either either the the nozzle nozzle basebase material material or or the the partial penetration welds partial penetration used in welds used in the the attachment to attachment to the the RPV RPV head head and and thethe risk assessment with risk assessment with regard regard to to nozzle integrity over nozzle integrity over aa period of period time.'

of time. .

Using Using the Framatome ANP the Framatome assessment, the ANPassessment, the DBNPS DBNPS feelsfeels assured assured in in operating operating until until thethe next scheduled refueling outage. This is based on the worst case scenario that a visible next scheduled refueling outage. This is based on the worst case scenario that a visible nozzle nozzle axial crackleak axial crack developed immediately leak developed immediately after after start-up start-up from from 10RFOlORFO in in May May 1996, 1996.

and was from one of the 19 drives that could not be inspected in 1998 (llRFO) or and was from one of the 19 drives that could not be inspected in 1998 (I1 RFO) or the the 24 24 .

drives drives that could not that could not bebe inspected inspected in in 2000 (12RFO). The 2000 (12RFO). The Framatome Framatome ANP ANP assessment assessment concluded that concluded such a crack that such crack would would take take 3.5 to to 10 10 years years toto grow cir:cumferentially through grow circumferentially through walL The DBNPS has assumed the 3.5 year value since wall ~e DBNPS has assumed the 3.5 year value since 3.5 years is based upon multiple 3.5 years is based upon multiple crack sites crack sites merging consistent with together consistent merging together with that that which which waswas observed observed at Oconee 3.

at Oconee 3. This This results in the development of a worst case through results in the development of a worst case through wall circumferential crackwall circumferential crack development developmentby by November November 1999 1999 (May(May 19961996 plus plus 3.5 3.5 years).

years). TheThe Framatome Framatome ANP ANP assessment further concluded that this crack would be expected to take assessment further concluded that this crack would be expected to take an additional 44 an additional years years toto grow grow to to maximum maximum ASME ASME Code Code allowable crack size allowable crack size ofof 270 degrees. Continuing 270 degrees. Continuing to apply this to the DBNPS's worst case scenario results in the potential to to apply this to the DBNPS's worst case scenario results in the potential to reach reach aa maximum allowable crack maximum allowable crack size size onon one one ofofthe obscured CRDM the obscured CRDM nozzlesnozzles (from(from 19981998 andand 2000 inspections) by November 2003. Because this date is beyond the date for the 2000 inspections) by November 2003. Because this date is beyond the date for the planned planned March March 2002 2002 refueling outage, the refueling outage, the DBNPS DBNPS has has concluded concluded that tbat there thereis is reasonable reasonable NRC004-1161 NRC004 ..1161

-"'"'"--' - ... ---~ .... - ... ~-.,:.-:- ...

Docket Docket Number Number 50-346 50-34.6 License Number NPF-3 Serial Number 2735 Serial Number 2735 Attachment Attachment I1 Page 440f5 Page of55 assurance assurance that that DBNPS DBNPS will will continue continue to operate operate safely safely to to the start start of I3RFO, 13RFO. scheduled scheduled for March March 2002.

Finite Finite Element Element Gap Analysis Analysis As discussed discussed with the NRC NRC staff staff during a telephone telephone conference conferenceca11call on on October October 3,2001, 3, 2001.

the DBNPS DBNPS contracted contracted with Structural Integrity Associates with Stt"ucturaiIntegrity Associates (SIA)

(SIA) to perform perform a finite fmite element element analysis of the RPV RPV head penetrations and nozzles.

head penetrations nozzles. This analysis analysis was performed performed to verify verify that gaps would exist between between the the CRDM CIU)M nozzles and the RPV head during during normal operation. These gaps would would permit permit through-wall leakage from any through-waIl leakage any nozzle or or through-weld cracks through-weld cracks.inin the J-groove observed via J-groove weld to be observed via boric acid crystal deposits.

~cid crystal This plant-specific plant-specific stress strc:Ss analysis used the DBNPS as-built analysis used as-built nozzle andand RPV head RPVhead dimensions. The analysis does not notinclude

.include the effects of primary primary system system pressure in .thethe nozzle gap area that would tend to further open the gaps. The SIAanalysis SIA analysis is included included herein as Attachment Attachment Sand 5 and provides assurance assurance that leakage leakage will be visible visible on all but four (4) of the sixty-nine (69) nozzle/penetration interfaces.

(69)nozzlelpenetration interfaces. However, the four four nozzle/penetration interfaces where it could nozzlelpenetration interfaces could not be be ~sured assured that leakage would be visiblevisible are nozzle numbers numbers 1, 1.2.2, 3, 3, and 4,

4. which are in the center center of ofthe RPV RPVhead.head. As documented documented in in the industry history history of circumferential cin::umferential cracks observed to date.

cracksobs'crved date, no DO leakage leakage attributable to circumferential attributable circumferential cracks has been been observed observed in this area from any of the inspections conducted inspections Therefore. based on the verification conducted by other licensees. Therefore, verification ofof inspection conducted at DBNPS.

inspection results conducted DBNPS, industry historical results of CDRM nozzle nozzle leakage and the finite element analysis penonned.

leakage the CRDM weld CRDM nozzlelhead nozzlelhead interface interface has previously cracking was not presenL weld cracking was not present previously occurred concluded that no leakage from performed, it is concluded occurred at the DBNPS.

DBNPS, and through- 0 INDUSTRY INDOS1RY EXPERIENCE EXPERIENCE & & FINDINGS Since discovery of Alloy 600 cracking at VC .Summer Summer and ONS.the ONS, the DBNPS has been following activities and pIanningsite-specific planning site-specific activities to assure that the ReactorCoolant Reactor Coolant System pressure boundary Systempressure boundary integrity is maintained. These activities have included included participation in industry groups that are extensively participation and characterizing the extensively analyzing andchatacterizing phenomenological attributes of phenomenological of the cracking issue, and developing sophisticated means of of detecting and, as necessary.

necessary, repairing identified identified cracks. The fmdings findings at other plants are being communicated communicated among the industry inna ina timely manner which which allows aggxessive aggressive evaluation of the nature, extensiveness andimplicatioDS nature. extensiveness and implications of the cracking to ensure the issue is understood as completely as possible, and ensures the development of conservative decision-making. his conservative It is through these continuing efforts as well as ongoing ongoing plant-specific plant-specific efforts that the DBNPS can also conclude that there is reasonable reasonable assurance assurance that the DBNPS will operate safely to its next refueling outage, scheduled to commence refueling outage. commence in March 2002.

NRC004-1162 NRC004-1162

Docket Number Docket Number 50-346 50-346 License Number NPF-3 License Number NPF-3 Serial Number 2735 Serial Number 2735 Attachment I1 Attachment Page 5 of Page of5 5 ALARA ISSUES ALARA In NRC 2001-01. page 8,Bulletin 2001-01, NRCBuiletin 8. the theNRe identified that nozzle penetration NRC identified activities penetration activities have the the potential personnel exposure.

potential for large personnel exposure. Plants have experienced experienced 15 to 40 rem rem CRDM nozzle during recent CRDM nozzle activities. bulletin states that all activities. The bulletin activities related to the alI activities the nozzles should be inspection of nozzles inspection be planned keep personnel implemented to keep planned and implemented exposures as personneJexposures (ALARA). As discussed in its initial response achievable (ALARA).

reasonably achievable low as reasonably response to to the the DBNPS will perform bulletin, the qualified visual perform qualified inspections or appropriate visual inspections appropriate supplemental inspection of supplemental inspection CRDM nozzle penetrations of the CRDM during its refueling outage penetrations during outage commence in scheduled to commence scheduled in March Inspection of these penetrations March 2002. Inspection between now penetrations between March 2002, and March and then again duringdlc 2002. and outage would refueling outage during the refueling significantly increase would significantly increase continued safe operation of the DBNPS can be exposures. Since the continued personnel exposures.

the personnel be reasonably assured to the beginning reasonably beginning of ofthencxt refueling outage, the next refueling completing additional outage. completing additional consistent with ALARA inspections before then would not be consistent inspections principles.

ALARAprinciples.

CONCLUSION CONCLUSION previous inspections Based on the previous analyses that conducted, analyses inspections conducted. have been that have performed been perfo~

concerning crack concerning growth rates, crack growth industry evaluations cracking. and industry rates. the ability to identify cracking, evaluations findings, it is concluded that there is reasonable assurance and findings. DBNPS will assurance that the DBNPS continue to operate safely to the start 13RFO, scheduled in March start of 13RFO.scheduled March 2002.

\:.

NRC004-1163 NRC004-1163

OOCket5~~6 Docket 50-346 License NPF-3 Ucense Serial Number 2735 2735 Page 1I of2 of 2 Nozzle Quadrant 1996 Inspection Core Quadrant Inspection results 1998 Inspection Inspection results 2000 Inspection results No.

No.

z Locat.

Locar.

See Note 1.0 1.0 1 H8 H8 11 Flange FlangeLeak Evident Flange Leak Evident Flange 2 G7 4 Range RangeLeak Evident Flange Leak Evident Flange 3 G9 1 Flange Range Leak Evident Flange Leak Evident Flange 4 K9 2 Range Leak Evident Flange Leak Evident Flange t:vident 5 K7 3 Range Flange Leak Evident Flange Leak Evident 6 F8 1 1 Evident Flange Leak Evident Flange Leak Evident Evident 7 H10 H10 22 Flange Evident Range Leak Evident Flange Leak Evident Evident 88 L8 L6 3 Observed No Leak Observed No Leak Observed Observed 9 H6 H6 4 No Leak Observed Observed No Leak Observed 10 FS 4 No Leak Leak Observed Observed No Leak Observed 11 11 Flo F10 .1 Flange Leak Evident Range evident Flange Leak Evident Evident 12 12 L10 L10 2 No Leak Observed Observed No Leak Observed Observed 13 L6 33 No Leak Recorded Recorded No Leak Observed Observed 14 14 E7 4 Flange Range Leak Evident Leak Evident Flange .Leak Evident 15 E9 1i Range Leak Evident Range Flange Leak Evident Leak Evident 16 Gl1 G11 1 Range Range Leak Evident Range Leak Evident Flange Evident 17 K11 2 No Leak Observed No Leak Observed Observed 1B 18 M9 M9 2 No Leak Recorded Recorded No Leak Observed Leak Observed

.. 19 M7 3 No Leak Observed Observed No Leak Recorded Recorded 20 K5 *3 3 No Leak Observed No Leak Observed Leak .Observed 21 21 G5 4 No Leak Leak Observed No Leak Observed Observed 22 DB D8 1 ,~, Flange Range Leak Evident Flange Leak Evident Evident 23 H12 H12 2 No Leak Observed Observed No Leak Observed 24 N8 N8 3 No Leak Recorded No Leak Recorded Recorded 25 25 H4 H4 4 No Leak Recorded Recorded No Leak Observed 26 ES E5 4 No Leak Recorded Recorded No Leak Observed 27 Eli E11 1i Range Range Leak Leak Evident Flange Leak Evident Evident 28 Mll M11 2 No Leak Recorded Recorded No Leak Observed Observed 29 MS US 3 No leak Leak Recorded Recorded No Leak Observed Observed 106 06 4 No*Leak Observed No Leak Observed No Leak Observed Observed 30 D1O 31 DID D10 1 FRange Leak Evl~ent Flange Evident Flange Evident Flange Leak evident 32 F12 F12 12 Range Leak Evident F.lange Evident Flange Leak Evident 33 L12 L12 2 No Leak .Recorded Recorded No Leak Observed Observed 34 N10 N10 2 No Leak Recorded Recorded No Leak Observed Observed 35 NS NS 3 No Leak Recorded Recorded No Leak Recorded Recorded 36 L4 L4 3 No Leak Recorded No Leak Observed Observed 37 F4 4 No Leak Recorded Recorded No Leak Observed 38 07 07 4 No Leak Recorded leak Recorded Flange Leak Evident Flange 39 C9 C9 11 Flange Leak Range leak Evident Evident Flange Leak EvIdent 40 G13 G 3 1

11 Leak Evident Flange Leak Evident .. Flange Leak Evident Evident 41 41 K13 K13 22 No leak Leak Recorded No Leak Observed Observed 42 09 09 22 No Leak Recorded Recorded No Leak Recorded Recorded 43 07 33 No Leak Recorded Recorded No Leak Recorded 44 K3 33 No .Leak Leak Recorded Recorded No Leak Observed 45 G3 G3 44 No Leak Leak Recorded Recorded No Leak Leak Observed 46 I 04 04 44 No Leak Recorded Recorded No Leak Observed 47 012 012 11 Flange Leak Evi~ent Flange Evident . Range Evident lange Leak Evident

- U - E mmmi a NRC004-1164 NRC004-1164

Docket50-346 Docket 50-346 Ucense UcenseNPF-3NPF-3 Serial Number2735 SenalNumber 2735 .

Attachment22 Attachment Page 2 of 2 Page2of2 Nozzle Core Nozzle Quadrant 1996 Core Quadrant Inspectionresults 1996Inspection results 1998 Inspectionresults 1998Inspection results Inspectionresults 2000Inspection

.2000 I'esults No.

No. Locat-local.

4 N 2 No Lo 4848 N12 N12 22 leakRecorded NoLeak No Recorded NoLeak No Observed lea~Observed 4949 N4 N4 33 No No Leak leak Recorded Recorded No Leak Recorded No leakObserved No Leak Observed No Leak Observed 50 C5 4

.50 C5 4 No leak Recorded No Leak Observed 51 51 Cl C11 I1 Flange LeakEvident FlangeLeak Evident FlangeLeak Flange LeakEvident Evident 52 52 E13 Et3 11 No leakRecorded NoLeak Recorded Flange leakEvident Flange Leak Evident 53 53 M13 M13 2 2 No LeakRecorded NoLeak Recorded No LeakObserved NoLeak Observed 54 54 Ol 011 22 No leakRecorded No Leak Recorded No LeakObserved No Leak Observed 55 55 05 05 33 No leakRecorded NoLeak Recorded LeakRecorded NoLeak No Recorded 56 56 M3 M3 33 LeakRecorded NoLeak No Recorded No LeakObserved NoLeak Observed 57 57 E3 E3 44 No leakRecorded NoLeak Recorded LeakObserved No Leak No Observed 58 58 58 B8 11 No leakRecorded No Leak Recorded LeakEvident FlangeLeak Flange EvIdent

... LeakObserved NoLeak No 59 59 H14 H14 22 No leakRecorded NoLeak Recorded Observed 60 60 PS P8 33 LeakRecorded No Leak No Recorded No LeakRecorded NoLeak Recorded 61 61 H2 H2 44 leakRecorded NoLeak No Recorded No LeakObserved No Leak Observed 62 62 B6 B6 44 leakRecorded NoLeak No Recorded LeakObserved NoLeak No Observed 63 63 610 BtO 11. No leakRecorded NoLeak Recorded Flange Leak Evident Flange.Leak Evident 64 64 F14 F14 11 No leakRecorded No Leak Recorded leak Evident Flange Leak Flange Evident 65 65 L14 l14 22 No No Leak Recorded leak.Recorded leakObserved NoLeak No Observed 66 66 pta Plo .22 No leakRecorded No.Leak Recorded No leakRecorded No Leak Recorded 67 67 P6 P6 33 No LeakRecorded NoLeak Recorded leak Recorded NoLeak No Recorded 68 68 L2 l2 33 No Leak Recorded No leak Recorded No leak Observed No Leak Observed 69

.69 F2 F2 44 _No leakRecorded No Leak Recorded leakObserved NoLeak No Observed Filedas

" Filed leakageissues/nozzle hlRCSleakage as WRCS review Table issues!nozzler6Vl8w Table Notes:

Notes:

11 In In 1996 during 10 1996during RFo, the 10RFO, entire RPV the entire headwas RPV head was inspected.

inspected.

Since the Since videowas thevideo wasvoid void of orientation narration, headorientation of head specificnozzle each specific narration, each nozzle view could not be correlated.

view could not be correlated.

Bold letters indicate Bold letters leaking CRDM Indicateleaking boltingflanges CRCM bolting discovered and flanges discovered during 12 repaired during and repaired (April 2000).

RFO (April 12RFO 2000).

No Leak Observed =

= Visual Inspection Satisfactory, No Video No Leak Observed VISual InspeclionSatisfactory, No Video Record Required. Record Required.

No Recorded --

leakRecorded No Leak .,.*Nozzle inspection recorded Nozzle inspection on recorded on videotape videotape Italicized text Italicized indicates nozzles text indicates nozziesthatare expected to not expected that are not leakage due show leakage to show due to insufficient gap.

to insufficient gap.

NRC004-1165 NRC004-1165

Number 50-346 Docket Number50-346 Numbec NPF-3 License Number Serial Number Number 2735 3 Attachment Page 1 of I Pagelofl RPV Head Inspection Results 3 pages follow NRC004-1166 NRC004-1166

RPVHeadIlRFlifIsuoction fPyread 11RIOfnsHee/ionResults Results Affected area from leaking tlange(s)

-No

@- leakage identified No leakage identified .

o- Evaluated not to have sufficient gap to exhibitleakage 0 - Evaluated not to have sufficient gap to exhibit leakage 3 -Insufficient gap with leaking flange Insufficient gap with leaking flange o- obscured by 0 - Nozzle obscured Nozzle by boron boron

• - Nozzle obscured by Nozzle obscured leaking flange with leaking boron with by boron flange NRC004-1167 NRC004-1167

RPVPVflesd Head12RF2 12RIO fnSHeetionIHesulis OInsiction Results Affected area from leaking flange(s)

@- - No leakage identified No leakage identified 0o --Evaluated Evaluated not to have sufficient gap have sufficient exhibit leakage to exhibit gap to leakage

1. -Insufficient

• - gap with Insufficient gap leaking flange with leaking flange 0* - Nozzle obscured by Nozzle obscured by boron boron

• • -- Nozzle Nozzle obscured obscured byby boron boron with leaking flange with leaking flange NRC004-1168 NRC004-1168

YPV9esd ll& 12 1F5 hisectionRlsuts Affected area Affected area since 11 RFO from leaking from leaking flange(s) flange(s) 9-@ -NoNoleakage leakageidentified identified o* - Evaluated not to have

- Evaluated notto havesufficient sufficientgap gaptotoexhibit exhibitleakage leakage

1. -Insufficient

- Insufficient gap with leaking flange o e - -Nozzle gap with leaking flange obscured Nozzle obscuredbybyboron boron

• -* - Nozzle obscuredbybyboron

- Nozzleobscured leakingflange withleaking boronwith flange Newly affected, since I I RFO, by leaking flange(s)

• - Newly affected, since II RFO,by leaking flange(s)

NRC004-1169 NRC004-1169

Docket Number 50-346 Docket Number 50-346 License License Number NPF-3 Serial Serial Number Number 2735 2735 Attachment Page Page 1 of 1 FRA-ANP 51-502567-0 FRA-ANP 51-502567-01. "RV Head 1, "RV Nozzle and Head Nozzle Weld Safety and Weld Assessment"n Safety Assessment Summary Summary Description Description The attached attached document document FRA-ANP 51-5012567-01 is FRA-ANP51-5012567-01 a non-proprietary isa non-proprietary updated updated version of a previously previously proprietary FfI document proprietary FI document (51-5011603-01).

(51-5011603.,Ql). This document document is the primary primary basis document document for the DBNPS's assertion that it is acceptable acceptable for the plant to to continue to operate continue operate until its next scheduled refueling outage next scheduled outage scheduled scheduled to start in March March 2002.

The The most important important portions this document portions of this document are are Sections Sections 3 and and 4.

Fifty-six (56) pages folloW pagesJollow Exhibit 24 Page 1 of 15 NRC004-1170

\

A-~ d-i  ?-:"

. . . . .5P.~

The attached document FRA-ANP 51-5012567-non-proprietary updated version of a 01 is a non-proprietary previously proprietaryFTI"document previpusly proprietary FTI document (51-(51-5011603-01). This document is the primary 5011603-01).

basis document for Davis-Besse's assertion that itit is acceptable acceptable for the plant top continue to scheduled refueling outage operate until it's next scheduled outage scheduled to start in March 2002.

important portion of this document is*

The most important is section 4.

It should be noted that Davis-Besse has It has contracted with SIAandsubmitted contracted SIA and submitted to the" the NRC SIA's plant specific stress analysis analysis which closely follows the stress analysis performed performed by FTI for all B&W plants which is covered covered in section 3.

3.

NRC004-1171 NRC004-1171

20440-8 20440-8 (1/2001)

(112001)

~RAMATOME OFRAMATOME ANP ANP ENGINEERING INFORMATION ENGINEERING INFORMATION RECORD Document Document Identifier Identifier ~ --~125()7 5012567-0

--101, )

nUe Title RV HEAD NO~QWELD HEAD NUZAN[ ~~ SAFETY ASSESSMENT uSAFETY ASSESSMENT.

PRePARED PREPARED BY. BY: REVIEWED REVIEWED BY:

BY:

Name Name SEE BELOW SEE BELOW Name Name KE. MOORE MOORE Signature Signature Oate ____- -

Date Signature Signature ,ffR/JikJ,..... Date Date9-jlf-QI Technical Manager Technical Manager Statement:

Statement: Initials Initials M.k Reviewer is Independent Reviewer Independent Remarks:

Prepared Prepared by:

Ltfo S. Fyfitc.

Fyfitcd(

.~

A.D. Nana P7(?e1 I

?f'z7 )9* '?4)7ýOL 01 IC Seals 1(/,r1t!

D.E. Kilian / 1 .........""--"

It? ~jl21I,,'

R.S. Enzinna $A.

/AWeimer Weimer* ~ ~~#

2 ft'MArA Cfjz.a/- /

This document document provides a safety assessment of reactor vessel head nozzles nozzles and welds that could could potentially susceptible to PWSCC in B&W-design potentially be susceptible incorporates information B&W-design reactors. Revision 01 incorporates information observed at ONS-2, a risk assessment, and editorial changes.

observed Page _1._1 of ~ 56 NRC004-1172 NRC004-1172

I NON-PROPRIETARY -

.... NON-PROPRIETARY 51-5012567-01 51-5012567-01 RV RV Head Nozzle and Weld Safety Assessment Table of Contents Section Section Title Page 1.0 1.0 Purpose 4 2.0 Introduction Introduction 4 3.0 Stress Analysis Efforts 12 4.0 Flaw Growth fnto Into th~

the RV Head 15 5.0 Leakage .Assessment Assessment 16 6.0 Wastage Assessment 16 7.0 Loos~

Loose Parts Assessment 17 8.0 Safety Analysis Review 18 9.0 Risk Assessment Assessment 21 21 10.0 Summary and Conclusions 33 33 11.0 References References 37 37 Appendix Appendix A Leak Assessments Leak

.~.

49 FIgure Figure Title Page Page 1 Side View Schematic Schematic of B&W-Design Reactor Vessel Head, Head, 41 41 Nozzles, Thermocpuple CRDM Nozzles, Thermocouple Nozzles, and Insulation Insulation 2 Plenum Cover Assembly Plenum 42 33 Control Control Rod Spider Assembly 43 44 Rod Guide Brazement Control Rod Brazement Assembly 44 5 Event Tree Tree for Frequency Frequency of Core Core Damage Damage from Outside Outside 45 Diameter Diameter PWSCC B&WOG CRDM PWSCC in .B&WOG CRDM Nozzle Nozzle 66 Crack Growth Growth Rate Assumed Assumed in Monte Monte Carlo Simulation Simulation 46 7 Probability Probability of Through-Wall Through-Wall Crack versus TimeTime after Initiation Initiation of 47 Outside Diameter PWSCC Outside DiarneterPWSCC 88 Probability Probability of Net-Section Net-Section Failure versus Time after after Initiation Initiation of 48 Outside Diameter PWSCC Outside Diameter PWSCC Page 22 of Page of 56 56 NRC004-1173 NRC004-1173

• • • • ,NON-PROPRIETARY

• "*NON-PROPRIETARY**** - 51-5012567-01 51-5012567-01 RV RV Head Nozzleand Head Nozzle andWeld WeldSafety SafetyAssessment As.sessment Figure Title Title Pgee Page 55 Radial 55 A-1 A~ 1 Radial Clearance ClearanceforforCenter CenterNozzle Nozzle Radial 56 A-2 A-2 Radial Clearance ClearanceforforOutermost OutermostNozzle Nozzle 56 Page Page 33 of of 56 56 NRC004-1174 NRC004-1174

.*** NON-PROPRIETARY ****

`***NON-PROPRIETARY **** 51-5012567-01 51*5012567-01 RV Head NoZZleNozzle and Weld Safety.Assessment Safety Assessment 1.0 Puroose Purpose The purpose of this report Is to provide an an assessment that demonstrates demonstrates safe operation ofB&W-design of B&W-design nuclear steam supply systems with the potential for nuclear steam for primary water stress corrosion cracking (PWSCC) of reactor vessel (RV) (RV) head penetration nozzles. This document document addresses addresses the the. assumed presence of assumed presence PWSCC in either the nozzle nozzle base material or the partial penetration penetration (or "J-..J-groove")

groovej welds used in their attachment to the RV head. This safety assessment assessment applies to the RV heads for the following nuclear stations:

Planf' Planta Owner Owner f---

Davis-Besse (D-B)

Davls*Besse(D*B) First Arst Energy Energy Nuclear Company Nuclear Operating Company Oconee Nuclear Station Units 1,2.1,.2 and 3 Duke Energy Energy Corporation (ONS-1. -2, and -3)

(ONS-1, -3)

Arkansas Nuclear One Unit 1I (ANQ.1)

Arkansas Nuclear (ANO-1) Entergy Operations, Incorporated Operations. Incorporated Crystal River Unit 3 (CR-3) Florida Power Rorida Power Corporation Three Wile Island Three Mile Island Unit 1 (TM1-1)

(TMI-1) Exelon Exelon Corporation a Note: This group will subsequently be identified as the "B&WOG plants."

a Note: This group will subsequently be identified as the "B&WOG plants."

Drawing Drawing on the applicable applicable results presented presented in several B&WOG documents severalB&WOG documents and the results of additional additional stress, structural, structural, flaw tolerance and fracture mechanics mechanics analysis, the objective of this document document is met met In addition, the results of aa review of the existing safety safety analyses (Section 8) shows that defense defense in in depth is assured. . .

2.0 Introduction Introduction Cracking was first observed observed in aa CRDM nozzle nozzle at the French pressurized water pressurized water reactor (PWR)

(PWR),Bugey Bugey Unit 3 in in 1991.

1991. SinC$

Since that time, the U.S, U.S. nuclear nuclear industry has developed developed safety assessments (References 1-3) and several utilities have assessments (References have 56 Page 4 of 56 NRC004-1175 NRC004-1175

• ... NON-PROPRIETARY NON*PROPRIETARY **** .... 51-5012567-01 51.5012567-01 RV Head Nozzle and Head Nozzle Weld Safety and Weld Safety Assessment Assessment proactively inspected control proactively inspected control rod drive mechanism (CRDM) nozzles drive mechanism nozzles considered considered susceptible to to be susceptible to PWSCC.

On April 1, 1997, the the Nuclear Nuclear Regulatory Commission, (NRC)

Regulatory Commission (NRC) issued issued Generic Generic Letter 97-01 97-01 (Reference (Reference 4). The TheB&W Owners Group B&W Owners Group (B&WOG) submitted BAW-(8&WOG) submitted BAW-(Reference 5) in 2301 (Reference 2301 In response Generic Letter response to Generic 97'()1, which provided details Letter 97-01, details of an integrated integrated inspection plan to address inspection plan address thethe potential degradation of RV potential degradation RV head penetration penetration nozzles nozzles at 8&WOG plants. [It is noted at B&WOG noted that the the B&WOG B&WOG plants have have two types two types of RV head penetration RV head nozzles, which consist of CRDM penetration nozzles, nozzles at all CRDM nozzles the plants and thermocouple the thermocouple nozzles TMI-1 only.]

ONS-1a and TMI-1 nozzles at ONS-1a All B&W-design reaelors were 8&W-design reactors were designed, fabricated, fabricated, erected, constructed. tested, erected, constructed, tested.

and continue to be Inspected in compliance Inspected compliance with tOCFR50.55a (Reference 6). In 10CFR50.55a (Reference particular, the RV head penetration nozzles were head penetration designed, fabricated, and were designed, manufactured manufactured to have a low probability probability of abnormal abnormal leakage, leakage,of of rapidly rapidly propagating failure, and propagating gross rupture and of gross accordance with General rupture in accordance General Design Criterion 14 of Appendix A to 10CFR50. 10CFR50. The Alloy 600 material utilized for these material utilized these

• RV

.RV head penetration nozzles is an austenitic penetration nozzles austenitic material ductile and material that is very ductile requirements set forth in General meets the requirements Design Criterion General Design Criterion 31 of Appendix Appendix A to 1 OCFR50. Finally, accessibility 10CFR50. accessibility to the RV head is available to assess assess the the structural and structural and leak integrity of the RV leak tight integrity AV head penetration nozzles head penetration nozzles in compliance compliance with General General Design Criterion Appendix A to 10CFR50.

Criterion 32 of Appendix tOCFR50.

The discovery of the J-groove Alloy 182 weld cracking at ONS-1 and the circumferentially-orlented flaw indications circumferentially-oriented indications revealed ONS-2, and ANO-1 revealed at ONS-3, ON&2, introduced new have introduced concerns that must be addressed. This document newconcems document provides provides a bounding bounding safety assessment to address the potential severity of these concerns at ONS, ANO-1,ANO-1, and the other 8&WOG B&WOG plants.

2.1 Backaround

Background

evaluation (Reference 3) presented a stress analysis, B&WOG safety evaluation The 1993 B&WOG crack growth analysis, leakage leakage assessment, and wastage assessment for wastage assessment potential inside surface PWSCC of the B&W-design CRDM nozzles. Based on performed, it was concluded that the peak hoop the results of the stress analysis performed, hoop greater stresses are greater than axial stresses on the inside surface of the nozzle. Also, the maximum hoop stress is similar-for sirnilar-for both the center and peripheral nozzles.

Thus, ifif an inside surface surface crack were to develop in a CRDM nozzle due to PWSCC,f the cracks would mainly be axially oriented. It was conservatively PWSCC conservatively concluded that safe operation of the 8&W-ciesign concluded B&W-design plants will not be affected affected for at least six years (operating with adequate leakage to corrode the AV RV head), and that within this time, the leak will be detected during a walk-down inspection of aThe thermocouple nozzles were a The thennocouplenozzles were removed from removed from ONS*1 ONS-1 at EOC-19.

EOC-19.

Page 50f56 Page 5 of 56 NRC004-1176 NRC004-1176

. RV Head

• • • • NON-PROPRIETARY Head Nozzle NON-PROPRIETARY - &***

Nozzle and Weld Safety Assessment Safety Assessment 51-5012567-01 51-5012567-01 the RV head area. Thus, the potential for cracking cracking of CRDM CRDM nozzles does not present a near-term near-term safety concem.

concern.

The same nozzle containing aa through-wall crack nozzle containing crack at Bugey-3 Bugey-3 also also exhibited an indication indication of circumferential circumferential cracking on its outside surface. In this case, the the initiation and propagation propagation of the axial crack preceded preceded exposure of the outer surface of the nozzle above the weld in the annulus to leaking reactor coolant.

leaking reactor An addendum addendum to the B&WOG B&WOGsafety safety evaluation evaluation was prepared prepared to address this address this concern in December concern (Reference 7). It was concluded December 1993 (Reference concluded in this evaluation evaluation that ample leakage through the penetration ample leakage penetration would occur to allow detection. detection. In addition, the occurrence addition. occurrence of nozzle detachment is highly unlikely during the design nozzle detachment life of the B&WOG plants since actions would be taken taken to repair the nozzle prior to toa a nuclear safety concern concern existing.

During a CRDM nozzle nozzle Inspection at Ringhals Unit 2 in 1992, an indication indication was was detected in the nozzle-ta-vessel nozzle-to-vessel (J-groove) weld at one penetration. The The indication was not indicative of PWSCC; reither, rather, the indication indication was attributed attributed to a weld defect that occurred occurred during fabrication of the CRDM nozzle to the RV weld.

The B&WOG B&WOG took action to address this concern concern by acquiring additional data additional data from several several sources. First, the data data from Ringhals Units 2 and 4 and data from a cancelled cancelled Westinghouse Westinghouse reactor, Shearon Harris, were acquired from the the Westinghouse Owners Westinghouse Owners Group Group (WOG).

(WOG). Second, the B&WOG perforrned performed an inspection inspection of the RV head from from* Midland Unit 1, which was a cancelled nuclear nuclear station fabricated fabricated by B&W.

Another addendum to the B&WOGsafety Another addendum B&WOG safety evaluation prepared to analyze evaluation was prepared analyze (Reference 8). This evaluation these data (Reference evaluation included a statistical statistical review andand analysis of the J-groove J-groove weld inspection data and a stress analysis of the CRDM J-groove determine the minimum J-groove weld to determine minimum weld area that is required to meet the the American Mechanical Engineers American Society of Mechanical Engineers (ASME)*

(ASME) Boiler and Pressure Vessel (B&PV) CodeCode primary shear stress limits. It was shown in this report that the the maximum maximum areas of weld lack of fusion detected detected for the Midland Unit 1, Shearon Shearon Harris, and Ringhals Ringhals Unit 2RV 2 RV closure heads heads are well below below the ASME Code allowable weld structural allowable limits for Weld concluded that a large structural integrity. It was concluded margin large margin exists between between the statistical statistical bound of the total lack of weld fusion areas in the the Midland Unit 1 head and the ASME ASME Code allowable limits. Therefore, Therefore, thethe observed lack of fusion areas do not give rise to a safety observed concern.

safety concem.

In addition, Generic Generic letter Letter 97-01 requested a description description of resin intrusions that may have occurred occurred at the B&WOGplants.

B&WOG plants. The B&WOG B&WOG response (Reference(Reference 5) historical records regarding included a review of plant historical regarding sulfate excursions. Also, the results of primary water chemistrychemistry analysis at each of the B&WOG 'plants plants were reviewed reviewed for excursions excursions from out-of-specification out-of-specification conditions. Based on these data, it was concluded that the potential for intergranularintergranular attack (IGA) or Page 6 of 56 Page60f 56 NRC004-1177 NRC004-1177

..* NON-PROPRIETARY NON-PROPRIETARY .... ""* 51-5012567-01 RV Head No~leNozzle and Weld Safety Assessment stress corrosion cracking (sec) (SCC) of CRDM thermocouple nozzles was very CRDM and thermocouple

~~

low. .

2.2 B&WOG Plant Inspections Inspections All B&WOG PlantsPlants As described described in References References 3 and 5, leakage of B&W-design B&W-design flanges has has previously been experienced experienced at each each of the B&W-design B&W-design plants, and visual inspections inspections Ofof the RV head area have been been implemented implemented so that flange leaks can be identified and repaired as soon as possible. Primary water that exits from aa leaking flange quickly flashes to steam, leaving behind a "snow" leaking "snow" of boric acidacid crystals. Exposure of the RV head to dry boric acid crystals crystals from this type of leakage has not resulted leakage resulted in wastage of the RV head.

The B&WOG utilities have included included plans to visually inspect the CRDM nozzle CRDM nozzle determine if leakage area to determine leakage is observed observed on top of the RV head, which would would indicate through-waJlcracking through-wall cracking has occurred, occurred, during their outages.

outages. In addition, walk-down inspections inspections have have been been implemented implemented in response to NRC Generic Generic (Reference 9) at each of the Letter 88-05 (Reference B&WOG plants. The walk-down theB&WOG inspections include an enhanced visual inspection of the gasket inspections gasket area and RV head during during every refueling refueling outage (12-24 months).

months). The B&W closure head and and service structure structure design provides provides access for a visual or boroscopic examination examination of the CRDM nozzle area, since the insulation insulation is not resting on the BV RV head head (see Figure 1). IfIf any leaks or boric acid crystal deposits are noted during inspection of the RV head area,area. an evaluation of the source of the leak and the extent of any wastage performed. This program wastage is performed. program has shown to be effective, as evidenced at ONS and ANO-1.

evidenced ANO-1. These visual examinations examinations provide an acceptable level of quality acceptable quality and safety and are in accordance accordance with 10CFR50.55a 10CFR50.55a General Design and General Design Criteria Criteria 30 of Appendix Appendix A to 10 CFR50.

BAW-2301 (Reference BAW-2301 (Reference 5) also describes describes the ASME B&PV Code Code Section Section XI, XI, Article IWB tWB 2500 inspections performed by all the B&WOG B&WOG plants. In addition, addition. aa plant-specific inspection plant-specific inspection of a CRDM nozzle and a thermocouple thermocouple nozzle was was performed performed by TMI-1TMI-1 in 1982 1982 as a result of intergranular intergranular attack on the steam generator tubes.

generator Most Most recently, NRC Bulletin 2001-01, 2001-01, "Circumferential Cracking Cracking of Reactor Reactor Pressure Vessel Head Penetration Head Penetration Nozzles" (Reference (Reference 10) was issued, plant-specific infonnation requesting plant-specific information regarding regarding the structural structural integrity integrity of the RV head nozzles and extent of leakage and cracking that has been head nozztes been found to date.

Information was also requested Information regarding inspections and repairs that have been requested regarding completed completed and those planned in the future to satisfy regulatory requirements, and basis for concluding that those plans will ensure the ba.sis ensure compliance compliance with the the Page 7 of 56 NRC004-1178 NRC004-1178

• • • • NON-PROPRIETARY

• • • • NON-PROPRIETARY .**. **** 51-5012567-01 RV Head Nozzle Nozzle and Weld Safety Safety Assessment Assessment regulatory requirements. Each of the B&WOG applicable regulatory B&WOG member utilities utilities prepared prepared a response response that provides proVides this information (References information (References 11-15). A summary of the plant-specific inspections inspections is described described below.

Oconee Unit Oconee Unit2 Duke Energy Duke Energy volunteered to perform inspectionbb of all 69 CRDM perform an inspection CRDM nozzles (from the nozzle ID) at ONS-2. ONS-2, which was ranked as one of the B&WOG B&WOG plants plants potentially susceptible susceptible to PWSCC, in 1994. Allindieations identified at ONS-2 in All indications identifted 1994 were confined to nozzle number 23 and consisted of 20 indications indications predominantly axial in orientation. These indications predominantly indications were detected with an eddy current technique current technique and confirmed confirmed with dye penetrant testing. An ultrasonic ultrasonic technique could not size the indications em on nozzle nozzle number 23 because because they were too shallow; therefore, therefore. the depth depth was conservatively conservatively assumed to be 2 mm (0.079 (O.079 inch). These These Indications were subsequently subsequently Identified identified as "craze-type" flaw indications. The 1994 eddy current results of a group of eleven indications. eleven other nozzles nozzles (numbers 16.' 16,-45,

45. 46, 50, 52, 56, 57.57, 60, 63. 63, and 65) indicated indicated high noise areas, with noZZle nozzle number 63 exhibiting the most severe noise of the group. Both Both ultrasonic and dye penetrant examinations were completed onnozzfe on nozzle number 63 without identifying -any any indications. BasedBased on this additional information, this this group of nozzles with high noisa noise was collectively collectively disposition dispositioned as non-reportable edas non-reportable indications.

indications.

.~,

Both rotating eddy current current and dye penetrant examinations examinations were completed on on nozzle numbers numbers 23 and 63 during .the the re-inspection work at ONS-2 in 1996.0 1996.c Multiple indications were observed (i.e (i.e.,.* small craZe-type craze-type flaw indications) that were predominantly predominantly axial in nature. Separate eddy current data acquisitions acquisitions were completed on both nozzles nozzles before and after use of a honing cleaning cleaning technique evaluate the effect of the cleaning on the eddy current results. They technique to evaluate were confirmed to be in the same location as the high noise areas areas detected with with eddy current in nozzle number 63 in 1994, 1994, thus explaining the cause of the noise noise that was previously previously unknown.

unknown.

In 1999, rotating eddyeddy current inspection results for nozzle numbers numbers 23 and 63 at ONS-2 again showed no significant significant change when the data were compared.to compared to the the 1994 1994 and 19961996 results. Thus it was concluded concluded that the indications indications had not grown or changed since the 1994 1994 inspection. Rotating eddy current results in 1999 1999 on nozzle numbers 16, 21, 21, 46, 50, 62, 50,62. and 68 were also obtained obtained and evaluated evaluated against the 1994 1994 results. As with all the previous data. data comparisons, no no significant change in the data was observed significant observed when compared compared to the 1994 data.

bNon-destructive examination techniques were developed and initially qualified for ID inspection 1/ Non-destructive examination techniques were developed and initially qualified for 10 Inspection of CRDM nozzles.

c Significant development work was completed to improve both the eddy current and liquid C Significant development work was completed to improve both the eddy current and Uquid penetrant penetrant methodologies.

methodologies.

Page 8 of 56 56 NRC004-1179 NRC004-1179

a*** NON-PROPRIETARY""

NON-PROPRIETARY **** 51-5012567-01 RV Head Nozzle and Weld Safety Assessment Head Nozzle Assessrnent Thus, it was concluded that the indications in these nozzles had not grown or changed since the 1994 inspection.

inspection.

recently performed ONS-2 most recently performed a routine visual inspection of the RV head during during 2001. Boric acid crystals were observed at four CRDM a refueling outage in April 2001.

nozzles (numbers 4, 6, 18, and 30). Liquid penetrant examination Uquid penetrant examination identified OD identified 00 crack-like crack-like axial indications below the weld on all four nozzles. nozzles. Ultrasonic Ultrasonic examinations examinations showed that these indications indications were OD-initiated OD-initiated and that none of the indications indications were through-wall.

through-wall. An OD-initiated circumferential indication, OD-initiated circumferential 0.07 indication,O.07 inch in depth 1.25 inch in length (approximately depth and 1.2Sinch (approximately 36 degrees circumferential extent),

extent). was noted above the weld on nozzle number 18 (Reference (Reference 11). Eddy examinations of the 10 current examinations ID of the nozzles revealed shallow shallow craze-type craze-type flaw clusters in all four fournozzJes

-nozzles that were distributed around around the entireentire 10 ID circumference (i.e., 3600, circumference 360°. above the weld). Based on these results, the leak path weld). Based was through the interface interface between between the nozzle and the J-groove J-groove weld.

The repair at ONS-2 automated repair.

ONS-2 consisted of an automated repair. The four CRDM nozzles CRDM nozzles were roll-expanded roll-expanded in the upper upper portion of the RV head in the area of the repair.

The bottom portions of the CRDM nozzles were were machined out to an elevation elevation above the original structural weld. The machinedRV machined RV headhead bores in the area for the new weld and the weld preparationspreparations in the CRDM nozzles were liquid liquid penetrant penetrant examined to verify that there were no rejectable indications indications in these these areas. The CRDM nozzles nozzles were weldedcto welded. to the RV head in accordance accordance with the the ASME Code using the temper bead technique technique using Alloy 52 weld material.

material. The The welds were then liquid penetrant and ultrasonic examined. The final operation fjnaloperation was to perform perform an abrasive water jet remediation remediation of the rolled and welded regions.

Oconee UnitUnit 1 As part of a routine visual inspection inspection of the RV head during during the ONS-1 ON8-1 refueling refueling outage (November 2000).2000), boric acid crystals were observed observed at one CRDM nozzle nozzle location (number 21) location 21) and at five of the eight thermocouple nozzle locations.

Eddy current examination of the inside surfaces of the thermocouple thermocouple nozzles nozzles showed showed that all eight nozzles contained crack-ike contained crack..:tike indications and that these were predominantly orientation. Ultrasonic predominantly axial in orieritation. Ultrasonic examinations examinations from the inside inside surface thermocouple nozzles allowed the weld size to be determined surface of the thermocouple determined .and and crack-like indications the axial crack-like Liquid penetrant indications to be located. Uquid penetrant examination of the the J-groove welds (after J-groovewelds nozzles), showed that some cracks had

{after boring out the nozzles).

penetrated through penetrated through the nozzle nozzle walls, and that the orientation orientation of these cracks was was predominantly axial at the plane where the cracks penetrated predominantly penetrated into the welds. All eight (8) of these nozzlesnozzles have been removed by sealing the RV head penetration with a more penetration more corrosion resistant material material (Reference (Reference 11).

11).

Page 9 Page of 56 9 of 56 NRC004-1180 NRC004-1180

A-" NONMPROPRIETARY NON-PROPRIETARY .... *"** 51 5012567-01 51-5012567-01 M

RV Head Head Nozzle Nozzle and Weld Safety Assessment Assessment Eddy current examination examination of the inside surfacessurfaces of CROM CRDM nozzle nozzle 21 and seven seven other other locations (42, 49, 55, 56, 61, 61, 67, and 68) was performed.performed. All eight of the the CRDM nozzles contained CROM contained craze-type indications located located in clusters in In the uphill region, both above and below the weld. Ultrasonic region,both examinations were performed Ultrasonic examinations performed on the inside surface surface of 18 nozzles (numbers 17, 21,22,28,34,42,47,48, 18 nozzles 21, 22, 28, 34, 42, 47, 48, 49, 52.54,55, 56,61, 52, 54, 55, 56, 61, 62, 66,67, 66, 67, and 68). No crack-like indications were detected.

A liquid liquid penetrant penetrant examination examination was performed performed on the partlat partial penetration penetration weld of nozzle 21.

. nozzle 21. Two Code acceptable acceptable small rounded indications were found. AfterAfter lightly grinding and performing performing another another penetrant examination, aa 0.75 Inch radial penetrant examination, indication indication running at a slightly skewed skewed angle angle across across the fillet weld was identified.

identified.

This crack was ground ground out of the weld and nozzle material. material. It extended extended into thethe nozzle material nozzle material approximately approximately 0.4 inch and ran radially radially out from the nozzlenozzle penetrating through the weld and through the butter layer in one location. This penetrating This crack was identified as the leak source since the annulus was exposed prior to to the crack being the crack fully removed.

being fully removed.

In summary, cracking was identifiedidentified in the CROM CRDM nozzle J-grooveJ-groove weld and continued from the weld into the 00 OD of the nozzle. CrackingCracking was also identified identified in the weld and nozzle nozzle 10ID of the eight thermocouple thermocouple nozzles. The cracking cracking mechanism was attributed to PWSCC.

mechanism PWSCC.AII All indications indications at these nine (9) locations locations were removed and weld repairs were performed.

Oconee Unit Unit 3 .,.

As part of a routine visual inspection of the RV head during an ON8-3 inspection of.the ONS-3 outage outage (February 2001\,

20011, boric acid crystals were were observed observed at nine CROM CRDM nozzle nozzle locations (numbers 3, 7, 11, locations 11, 23.

23, 28, 34, 50, 56, and 63). In addition to observations of through-wall axial flaws above the weld, outside surface observations surface circumferential indications (relatively deep and located circumferential located below the weld) were present present on four four nozzles (numbers 11, 11, 23.

23, 50, and 56). Also, outside surface surface circumferential indications circumferential indications above above the weld (one through-wall and one nearly through-wall) were through-wall) were present on two nozzles nozzles (numbers $0 50 and 56).

56). Another Another surface circumferential outside surface circumferential indication indicatior:l above the weld, which was relatively relatively shallow (0.22 inch deep),

deep), was present on nozzfe nozzle number 23. It appeared that these particular cracks cracks initiated from the nozzle nozzle OD00 following exposure to I$akingleaking primary primary water from a through-wall axial flaw (Similar (similar to the Bugey-3 Bugey-3 cracking).

would require an axial crack, and ultimately a leak path.

This wouJdrequire path, to either either propagate propagate through the weld or the nozzle surface surface adjacent adjacent to the weld..weld. The fact that nozzles nozzles 50 and 23 had a circumferential circumferential crack crack at the OD 00 that had not propagated propagated through-wall supports through-wall supports the assertion that an axial needed prior to aa axial flaw is needed circumferential circumferential flaw being formed. The existence of seven seven (7)(7) other nozzles 'With with at least one axial indication indication connected connected to the 00 OD surface nozzle at ON8-3 surface of the nozzle ONS-3 also supports this position.

"Page 10 of 56 NRC004-1181

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• • NON-PROPRIETARY **** 51-5012567-01 51-5012567..()1 RV Head Head No~zle Nozzle and Weld Safety AssessmentAssessment penetrant examination of the weld excavation Liquid penetrant excavation area in nozzle number 34 also revealed a circumferentially circurnferentially oriented flaw Indication.

indication. Several PT examinations examinations performed during the excavation process excavation process revealed this that this particular .indlcation indication spanned about 2-inchesin 2-inches in length and was located located in the J-groove weld. This indication did not appear appear to extend to the reotof root of the weld.

Shallow axially oriented inside surface ShaIJowaxialJy surface indications were also observed observed In in areas of craze-type craze-type cracking above and below the weld (similar (sirriilar to those .observed observed at ONS':'1 and 2) in virtually all the nozzles ONS-1 nozzles examined.

examined.

The circumferential circumferential cracks discovered at ONS-3 ranged from about 2-rnm discovered atONS-3 2-mm (0.079 Inch)

Inch) in depth to through-wall.

through-wall. In In nozzle number 11. 11, a circumferentially clrcumferentially oriented oriented 00 crack (below the weld) crossed OD crossed the path of three axial cracks, the the circumferential crack circumferential crack was 0.380 inch deep (61% (61 % through-wall), with a 31 31% %

circumferential circumferential extent (Reference 11). In nozzle number (Reference 11). number 56, a circumferentially circumferentially oriented OD 00 crack (above the weld) welo) was through-wall through-wall and extended extended approximately approximately 165 16500 around the nozzle. The circumferential circumferential crack in nozzle nozzle number number 50 50 was nearly nearly through-wall (i.e., pinhole indications were observed observed on the 10 ID during during liquid penetrant penetrant testing) and extended approximately 59 extended approximately 5900 around around the nozzle. The circumferential circumferential extent and through-thiCkness through-thickness depth (from the OD) 00) of nozzle nozzle number 23 were 6s<' 66" around the nozzle nozzle and 0.22 inch inch (Reference (Reference 11).11 ).

All through-wall cracking AU cracking observed below the weld appears to have have initiated initiated near the toe of the fillet weld. It is most likely associated with the residual weld weld stresses introduced introduced in this areaarea during manufacturing. Some shallow shallow cracking was also observed observed on the outer surface surface at the end of sever~1 several nozzfes nozzles (e.g.,

number 28 and 56).

Ultrasonic Ultrasonic examinations subsequently performed examinations were subsequently performed on an additional set of nozzles (numbers 4, 8, 10, 14.

nine nozzles 14, 19.

19, 22,

22. 47, 64, and 65). 65). From these nine nine nozzles, eight of them showed no indications. indications. Nozzle number 4, however, Nozzle number showed four shallow axially oriented flaws, all on the inside surface surface and aboveabove the weld. Also, eddy current examinations examinations were performed for these additional nine nozzles. For nozzle nozzle number number 4, the eddy eddy current results confirmed the findings from the ultrasonic examination.

ultrasonic eXamination. In addition, the eddy eddy current examination revealed revealed shallow craze-type flaw.clusters shallow craze-type flaw clusters that were found in four nozzles (numbers 8, 10, 14, and 22) and distributed distributed around around the entire 10 ID circumference (360°,

circumference {360°, aboveabove the weld).

weld}.

Arkansas Nuclear Arkansas Nuclear One Unit 1 Following shutdown shutdown for a scheduled scheduled refueling outage (March 2001). 2001), ANO-1 AN0-1 performed aa routine visual inspection performed inspection of the RV head area. This inspection head area inspection revealed revealed boric acid crystals in the area. area of one CROM CRDM nozzle (number 56). 56). Based on the visual inspection results, liquid penetrant penetrant (PT).

(PT), ultrasonic (UT),(UT). and eddy current (ET) examinations examinations of the.CRDM nozzle and PT oftheJ-groove of the J-groove weld weJdwere were Page 11 of 56 56 NRC004-1182 NRC004-1182

H** NON-PROPRIET NON-PROPRIETARY AAY*--- 51-5012567-01 51-5012567-01 RV Head Nozzle Nozzle and Weld SafetySafety Assessment Assessment also performed.

performed. The leak path was determined to to be an axial flaw in the nozzle nozzle outside diameter that extended diameter that$xtendedbeyondeach beyond each side of the the weld. The PT examination identified identified a circumferential circumferential crack, approximately approximately 0.70 inch long in the the outside diameter diameter of the CRDM nozzle below the J-groove weld (Reference (Reference 12).

This crack branched twice and each of the three resulting tributarieS crack branched tributaries extended off-axial (nearly axial) up to the weld fusion line. .Iine. There is no firm evidence evidence that any cracking occurredoccurred in the weld. The UT examination examination indicated indicated that the the subsurface subsurface dimensions dimensions of the crack extended in a circumferential circumferential and then off-axial direction direction below the weld and in an axial direction direction thrO!Jgh through the nozzle past the weld to a termination point 1.3 inches inches above the weld on the nozzle nozzle 00 OD On (in the annulus region). The flaw depth dimension was estimated to be be a maximum maximum of 0.20 inch into the nozzle wall and thus never penetrating penetrating to the nozzle nozzle 10 ID surface.

surface. It would appear from the NDE evidence that the cracking

.the cracking was confined confined to the nozzle material, material, which became became the leak path. The TheET ET and PT examination examination confirmed that the crack had not propagated confirmed propagated to the inside diameter diameter of the CRDM nozzle. This flaw is consistent with the PWSCC experience experience that has occurred at at ONS-1.

ONS-1. This event also reaffirmed reaffirmed the effectiveness effectiveness of examining the RV closure closure head for leaks as means of assuring the avoidance avoidance of a safety safety concern.

3.0 Stress Analysis Efforts Efforts 3.1 Summary Summary of Stress Analyses Performed Performed Nonlinear elastic-plastic Nonlinear elastic-plastic finite element analysis was performed eiementanalysis performed in 1993 to characterize stresses in the Alloy 600 CRDM characterize CRDM nozzle (SB-167 (88-167 tube material),

material), the tile low alloy ste.J lowalloyste ...J head, the stainless stainless steel cladding in the head, and the Alloy 182 182 weld material used for the partial penetration weld and butter between between the nozzle nozzle and head. The purpose of this analysis was to determine d.eterrnine the preferential direction for cracking based on the relative magnitude magnitude of longitudinal longitudinal (axial) andand circumferential (hoop) circumferential (hoop) stresses. Results were also used to predict predict crack growth by PWSCC and to to support support leakage assessments assessments for postulated postulated through-wall through-wall cracks in the nozzle wall (Reference (Reference 3).

Two bounding nozzle configurations were addressed nozzle configurations addressed in the 1993 stress analYSiS, analysis, nozzle and one of the outermost the center nozzle outermost peripheral nozzles (hillside nozzle).

peripheral nozzles Taking advantage advantage of full symmetry of the top of the reactor reactor vessel vessel head, the the center nozzle was analyzed analyzed using a two-dimensional model.

two-dimensional model. Since Since the outer hillside nozzle penetrates hillside nozzle penetrates the head head at an angle of 38.5 degrees, a 180 degree degree three-dimensional three-dimensional model was utilized utilized at this location to address the more more complicated complicated stressstress fields associated associated with an oblique penetration, p.enetration, due in part part to ovalization of the nozzle nozzle under under pressure and thermal loads. The following following loading conditions loading conditions were considered in the 1993 stress analysis analysis to determine determine long-term sustained stress in the nozzle and weld materials (and to a lesser long-term extent in the head):

Page 12 of 56 NRC004-1183 NRC004-1183

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1) Shrink fit of the nozzle within the head during installation (0.0010 inch nozzle within' inch nominal diametric diametric interference).

interference).

2) Simulated Simulated welding welding of the nozzle to the head (heatup of the weld material to 2470 OFof and cooldown to develop residual stresses).

3) Cold hydrostatic hydrostatic testing of the completed head assembly at a pressure of 3125 psig.

4) Steady Steady state operation a temperature of 600 of operation at atemperature OF and a pressure pressure of 2250 psig.

Residual stresses from the welding process dependent on plastic process are strongly dependent plastic deformation deformation in the nozzle.

noZzle. Yield strengths for B&W-design B&W-design plants range from 31 31 ksi to 64 ks!.

ksL. For the higher yield strength nozzles, more more residual stress is locked in as the weld puddle cools from its molten molten state. The 1993 1993 .stress stress analysis used the 64 ksi nozzle nozzle yield strength as a bounding value.

3.2 Nozzle Nozzle and Weld Stresses Stresses Since at most locations the inside surface surface hoop hoop stress is higher than the axial stress, the preferential preferential direction direction for cracking is axial (in a radial plane relative to nozzle). Exceptions the nozzle). Exceptions occur occur at the lower end of the nozzle nozzle and above the weld.

Some 'circumferential cracking may occur circumferential cracking occur on the outside outside surface surface of the nozzle, just below the weld, where hoop and axial stresses are similar in magnitude magnitude on the uphill side. Axial stresses would also promote the propagation propagation of OD00 initiated initiated circumferential cracks above the weld.

3.3 Evaluations Flaw Growth Evaluations Evaluations Evaluations of flaw growth from PWSCC fJawgrowth PWSCC have been performed performed for the J-groove J-groove weld and CRDM nozzle as discussed below. Axial ID weld 10 nozzle nozzle flaws were were addressed in the original safety addressed evaluation for cracking safety evaJuationfor B&W-design CRDM cracking of B&W-design nozzles (Reference 3).

nozzles (Reference 3.3.1 Axial J-Groove J-Groove Weld and 00 OD CRDM Nozzle Nozzle Flaws As discussed above, the dominant dominant hoop stress in the J-grooveJ-groove weld would would promote promote axial cracking material. Due to the relatively high crack cracking of this Alloy 182 material.

growth observed in autoclave growth rates observed autoclave tests with this weld metal in aa PWR PWR environment environment (Reference 16), 16), and considering the increasing increasing stress gradient gradient away from the inside inside surface surface of the weld, crack growth through the J-grooveJ-groove weld weld would be expected. Although Although the flaw would arrest arrest at the low alloy steel RV Page 13 of 56 Page NRC004-1184 NRC004-1184

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head 4.0), the flaw would continue to grow into the Alloy 600 GRDM CRDM nozzle, as seen at ON5-1 nozzle. ONS-1 (Reference 17).

Calculations performed to predict the time it would take to grow an axial Calculations were performed outside surface surface flaw (OD flaw) through the no~eto (00flaw) nozzle to the Inside surface.

Assuming a length-to-depth length-to-depth ratio of six, using using the Peter Scott crack growthgrowth model for Alloy 600 ina in a PWR environment, and considering the highest highest stressed location, it would take almost four years for an axial OD flaw that is initially 0.5 mm (0.02 inch) deep to grow through-wall. It has already been reported (Reference 3) that it would take at least four more years for a through-wall through-wall flaw to extend two Inches extend above the weld, thereby inches above thereby creating creating a leak leak path into the annular annular region between between the nozzle and head.

Circumferential 00 3.3.2CircumferentiaJ 3.3.2 OD CROM CRDM Flaws Since Since the 00 OD surface surface hoop stresses in the weld are about about two times the surface surface axial stresses; stresses; flaws originating at this location should be oriented oriented in an axial plane. Development Development of ofa a leak path through the weldw~ld to the annulus between between the nozzle nozzle and RV head would however, expose however,expose the outside surface of the nozzle nozzle to the primary water environment. Since there is a band band of high axial stress on the the outside of the nozzle just above above the weld, initiation initiation of aaclrcumferentiaJ circumferential crack at this location is aa concern. experience at concern. Based on elglerience ONS-3, the development atONS-3,the development of an axial leak path throughthrough the weld andlorand/or nozzle would precede initiation initiation of a circumferential OD circumferential 00 flaw on the outside surface surface of .the the nozzle above the weld.

Furthermore, as observed at ANO-1, Furthermore, ANO-1, deposits of boric add acid crystals on the top of the head would provide evidence evidence of aa leak path prior to the initiation of a circumferential circumferential 00 OD flaw. For the purpose of performing crack growth calculations, it is conservatively calculations, assumed that a small flaw, 0.5 mm (0.02 inch) conservatively assumed inch) in depth, initiates immediately depth,initiates immediately after the plant is returned to service. Using Using 0.5 mm mm (0.02 inch) as the initial initial depth of an isolated OD circumferential flaw 00 initiated circumferential above the weld, it would take more than 10 years for aa short (Va = 6) semi- =

flaw to grow through-wall.

elliptical surface *flaw through-wall. At ONS-3, following following the growth of an axial axial flaw to the annulus between between the nozzle and head, there were apparently apparently several several initiation sites that linked to form a long circumferentialcircumferential ODoutside OD outside surface surface crack above the weld, extending nearly half way around the the circumference.

circumference. Such a flaw could grow grow through-wall in 3.5 years. Even then, it would take another another 4 years for the through-wall flaw to grow another another 25% aroundaround the circumference.

circumference. The remaining ligament, which would then be 25% of the the original circumference, circumference, would still be sufficient sufficient to preclude gross net-section net-section failure (nozzle ejection).

ejection). This ligament ligament satisfies primary stress limits using a safety safety factor of 3 (Reference (Reference 18).18).

Lack of fusion weld defects between between the nozzle and weld, weld, of the type detected detected at Ringhals Unit 2 and at the cancelledcancelled Shearon Harris and MidlandMidland plants, should should Page 14 of 56 NRC004-1185 NRC004-1185

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.... NON-PROPRIETARY ARV -**** 51-5012567-01 51-5012567-01 RVH,~d RV Nozzle and Weld Safety Head Nozzle Safety Assessment also be considered considered in light of th& the potential for CRDM nozzle J-groove J-groove weld cracking. This flaw is described circumferential flaw along described as a "wrap-around" circumferential the cylindrical surface at the nOzzle-to-weld nozzle-to-weld interface.

Interface. As discussed in Reference Reference 8, there may be up to 67% lack of fusion between between the nozzle nozzle and weld before the the ASME Code primary shear stress limits are violated. It has been calculated calculated that it would take two years for a O.25-:inch 0.25-inch wrap-around wrap-around flaw to grow to the6"TOk the 67% limit.

This is based ona on a conservative conservative value of 45 ksi ksl for the average radial stress stress between between the nozzle nozzle and weld, and utilizes the high crack growth rates observed observed in laboratory laboratory testing for Alloy 182 weld metal (Reference 16).

metal (Reference 16). Based on observations observations at ONS-l,ONS-1, ONS-2,* ON~3, and ANO-1, ONS-2, ONS-3, ANO-l, where there was no no evidence evidence of wrap-around wrap-around cracking cracking between between the nozzle and weld. weld, this is an extremely conservative crack growth prediction.

extremely conservative A 2-inch circumferentially oriented flaw indication was observed 2-inch long circumferentiallyorientedflaw nozzle 34 observed in nozzle at ONS-3.

ONS-3. It was located in the weld material material and spiraled spiraled from a distance  %

distance of 111/8 inch from the 00 OD of the nozzle on the uphill side to 0.75 inch from the nozzle nozzle as it went about 45° 450 around around the weld. Being located located in the weld, this laminar-type laminar-type anomaly anomaly is not considered to be a safety concern, since it did not provide provide a leak path to the environment and it could not lead to ejection of the nozzle.. nozzle.

4.0 Flaw Growth Growth Into the RV HeadHead A crack, propagating J-groo*,e weld by pwsce.

propagating through the J-groove PWSCC, will eventually grow to the RV head (low alloy steel) and the CROM CRDM nozzle nozzle (Alloy 600). It is expected expected that the resultant crack will continue to propagate through the CRDM eROM nozzle nozzle material material as observed at ONS-1 (Reference (Reference 17) and ONS-3, In a direction direction determined determined by the residual stress distribution. However, continued flaw growth into the low alloy steel is not expected expected to occur.

Stress corrosion corrosion cracking (SCC) of carbon and low alloy steels is not expected cracking (SeC) expected to be a problem under BWR or PWR PWR conditions (Reference 19). SCC steels sec of steels containing up to 5% chromium is most frequently frequently observed in caustic and nitratenitrate solutions and in media media containing containing hydrogen hydrogen sulfide (References (References 20 and 21). 21 ). A recent recent review review of literature results was performed performed by Framatome Framatome ANP, which also concluded that sec concluded SCC of low alloy steel materials materials is non-credible non-credible in PWR PWR environments environments (Reference information, sec (Reference 22). Based on this information, SCC Is not expected expected to be a concern for low alloy steel exposed exposed to primary primary water.

Instead, Instead, an interdendritic interdendritic crack propagating from the J-groove J-groove weld area is expected expected to blunt and cease propagation. This has been shown shown to be the casecase interdendritic sec for interdendritic SOC of stainless steel cladding cracks cracks in charging pumpspumps (References 23 and 24) and by recent events events with PWSCCof PWSCC of Alloy 600 weld weld materials materials at ONS-1 ON5-1 and VC Summer (References (References 25 and 26). Although a PWSCC-initiated PWSCC-initiatedflaw flaw may continue to propagate by fatigue crack growth into the the Page 15 of 56 NRC004-1186 NRC004-1186

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--*NON-PROPRIETARY w** 51-501258T-01 51-5012561-01 Head Nozzle RV Head Nozzle and Weld Weld Safety Safety Assessment Assessment low alloy alloy steel con.sidered to head, this is considered steel head, Insignificant over to be insignificant several operating over several operating cycles based cycles based on loads. Since anticipated cyclic loads.

on anticipated borated water Since borated water will now now be be in in contact contact with wastage of alloy steel, corrosion wastage with the low alloy material is expected of the material expected to occur. This Is is addressed addressed in Section 6.0 below.

Section 6.0 5.0 Leakage Assessment Leakage Assessment B&WOG has The 8&WOG performed leakage has performed leakage assessments potential leak assessments for various potential leak scenarios expected prior to the recent leak scenarios expected leak events ONS-1, ONS-2, ONS-3, and events at ONS-1, ONS-2, and ANO-1. The results ANO-1. results from these these assessments documented in detail in assessments are documented Appendix A. The recent experience, Appendix leak rates are indicates that the leak experience, however, indicates are apparently very low based apparently based on the amount amount of boric acid crystals observed boric acid observed on leaking nozzles. It was estimated leaking nozzles. approximately 0.5 in3 was present estimated that approximately present around around CRDM nozzle CRDM number 21 at ONS-1.

nozzle number ONS-1. In the case case of the ONS-1 thermocouple ONS-1 thermocouple nozzles, five (5) were suspected suspected to have have leaks while the otherother three (3) did not three (3) evidence of boric exhibit evidence boric acid crystals. examinations subsequently The examinations subsequently performed on all eight performed eight (8) nozzles nozzles revealed cracking that would strongly suggest revealed cracking suggest reasoned that a small leak a leak path. It is reasoned narrow annulus can lead to "leak leak and narrow formation of less dense plugging" by the formation metal oxides dense metal annulus. Thermal oxides in the annulus.

anticipated to lead to starting or re-initiating cycling is anticipated weeping type re-initiating a weeping type leak.

Therefore, anticipated to be minimal until leakage is anticipated Therefore, leakage until a long axial flaw (i.e.,

approximately approximately the the length length ofof the the RVRV head penetration) develops head penetration) above the develops above the weld.

weld.

6.0 Wastage Assessment Wastage Assessment The purpose of this section Is potential damage that can occur is to assess the potential occur to the RV head head as a result of a leaking leaking CRDM nozzle nozzle or J-groove weld. Two areas considered In this discussion:

concern are considered of concem

1) General corrosion damage General reactor vessel head as a result of damage to the reactor exiting boric acid crystals and borated steam, steam, condensing on the insulation from a through-wall crack in a CRDM nozzle or J-head insulation groove weld.

2) Corrosion damage both within and in the vicinity of the reactor Corrosion corrosion resulting from a penetration due to boric acid corrosion vessel head penetration through-wall through-wall crack in the CRDM nozzle or J-groove J-groove weld.

A leaking CRDM nozzle or J-groove J-groove weld is of concem because the leaking concern because leaking primary coolant, containing boron in the form of boric acid, can be very corrosive corrosive to carbon and low low alloy steel materials subjected to certain environmental materials when subjected conditions. Several studies have have been performed performed to determine these conditions.

Page 16 of of 56 NRC004-1187 NRC004-1187

..... NON*PROPRIET NON-PROPRIETARY ARY ...- 51*5012567-01 51-5012567-01 Nozzle and Weld Safety RV Head Nozzle Safety Assessment Assessment description of the testing performed and their respective A description respective results is given in in References 3 and 27.

References damage assessment for a variety corrosion damage Reference 3 includes a corrosion Reference variety of conditions conditions leakage rates assumed to occur with CRDM nozzles. As noted above and leakage above in Section 5, similar assumptions can be Section be made for the case of leakage that is associated with PWSCC associated PWSCC of RVhead J-groove welds.

RV head J-groove ItIt was determined Reference 3 that this type of leakage would lead to determined in Reference corrosion of the RV head penetration, at corrosion a maximum volumetric metal loss rate of ata 1.07 in3/yr. Three defect profiles were postulated to model this level of corrosion concluded through an ASME B&PV for a time period of six years. It was concluded 8&PV Code membrane stresses In the RV head, that safe operation evaluation for membrane evaluation operation of thethe corrosion of the RV head affected as a result of this level of corrosion plant would not be affected penetration.

penetration.

B&W-design plants will not be concluded that safe operation of the B&W-design Finally, it was concluded be affected for at least six years, and that within this time, the leak will be detected affected for detected walk-down inspection of the RV head area. It should be noted that this during a walk-clown this minimum six-year period represents corrosion minimum corrosion of the RV head at the maximum maximum rate of 1.07 in3/yr, which would only occur sufficient leakage rate has occur when a sufficient has potential for cracking of CRDM nozzles been realized. Thus, the potential nozzles and RV head J-groove welds does not present a near-teml near-tewn safety safety concern. The validity of these assumptions and conclusions assumptions recently verified conclusions was recently verified by the detection of boric acid crystal deposits around CRDM and thermocouple nozzles and the subsequent thermocouple nozzles identification of RV t-~ad identification l-,ad J-groove J-groove weld leakage ONS-1, ONS-2, ONS-3, and leakage at ONS"1, and ANO-1. In all cases, only minimal corrosion (wastage)

ANO-1. (wastage) was observed.

7.0 Loose Assessment loose Parts Assessment As noted circumferential cracking has been noted earlier, circumferential been observed on the outsideoutside surface of leaking surface leaking CRDM nozzles nozzles at ONS-3. This cracking occurred at the toe of cracking occurred attachment to the RV the fillet weld that forms part of the structural attachment RV head. In In some of these nozzles through-wall through-wall axial cracking has also been been observed in the nozzle base metal belowbelow the weld. Thus, there is a concem Thus,there concern that a through-wall circumferential crack could link up with two or more through-wall axial cracks and circumferential part. An assessment of the potential form a loose part consequences associated with potential consequences fragmentation has been CRDM nozzle fragmentation been performed (Reference 28). The potential performed (Reference transport of fragments originating at the reactor transport reactor vessel head penetration penetration were identified and evaluated.

evaluated.

If a piece of the CRDM nozzle were to break away, it could potentially Ita potentially end up In one of three places. The first location is the stainless steel plate around the column weldments (plenum cover) where have an impact on any where Itit would not have Page Page 17 of 56 NRC004-1188 NRC004-1188

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. RV Head Nozzle Nozzle and Weld Safety Assessment Assessment safety or operational operational issue in the plant (see Figure 2). The second location is Is through the gaps aroundaround the periphery periphery of the plenum cover and would likely end up in in the steam generator, potentially damaging damaging the tubes or tube welds. A A fragment lodged within a single tube cot.dd, could, as a result of motion induced induced by thethe flow through the tube, cause cause wear wear of the tube at the point of contact contact with thethe inside surface. Although unlikely, this coufd could eventually eventually result in a small sma" through-wall flaw flawinin the tube, causing aprimary-to-secondary a primary-to-secondary leak. leak, which which can be detected detected by monitoring procedures already in place monitoring procedures place at the plant. Once detected, detected, the plant operators would follow the technical specification 0perators specification action action statements to shut down down if the leak became became significant. This does not introduce any new new or unanalyzed unanalyzed event. While this location may cause equipment equipment damage, it is not a safety concern. The third possibility.

concem. possibility, which could be asafetyconcem,is a safety concern, is that the pieces pieces could enter .anyany one of the 69 column weldments weldments through which whlcrt the control rod spiders descend descend (see Figures 5 and 6). 6). It has been calculated that there .is is a 25% chance 25% chance or greater greater for a loose loose piece to enter enter one of the column weldments.

This is simply based based on an area ratio of the column weldments in the upper head and the fact that low cross flow velocities in this region would tend to allow debris debris to fall vertically. In addition, the leadscrewscould leadscrews could tend to guide the debris debris such that the probability of entering entering the column column weldment may be be much higher than much higher 25%. If fragments enter the column weldments, they will likely be be stopped stopped on one of the control rod guide tube brazementsbrazements where relatively small fragments fragments

<<  % inch) would be

(< 34 capable of precluding becapable.of precluding complete control rod insertion.

experience at ONS-3, circumferential Based on experience circumferential and axial cracking below the the weld is accompanied by through-wall through-wall axial cracking cracking at and above the weld. The The experience coupled with the extensive examinations performed ONS experience performed in Europe, and the stress analysis results described described in Sf S, "tion

-tion 3.0 indicate that the the predominant cracking orientation is axial.

predominant In addition, addition, there have have been been a total of 27 non-leaking non-leaking nozzles at both ONS-1 ONS-1and and subjected to both eddy current and ultrasonic ONS-3 subjected ultrasonic examinations. Very shallow Veryshallow craze-type cracks were revealed above craze-type above and below the welds. No 00 OD cracks cracks detected at the were detected nozzle-to-weld intersection thenozzle-to-weld intersection (below the weld) for these 27 27 nozzles. In each case, these nozzles nozzles. nozzles were found to be free of cracking. These These observations and results support the assertion that there is a high probability that observations detectable leakage would precede the development detectable leakage development of a loose part.

8.0 Safety Analysis Review In this section, the plant safety analyses will be reviewed to determine determine if aa safety Issue exists and to provide justification that the consequences provide justification consequences of a failure of a single CRDM nozzle are bounded by the existing plant safety CRDM nozzle safety analyses analyses and will support plant restart and continued operation.

operation.

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• • NON-PROPRIETARY ..** 51-5012567-01 51-5012567-01 He~d Nozzle RV Head BV Nozzle andand Weld Safety Assessment Weld Safety Assessment Loss of Loss of coolant coolant accidents accidents (LOCA) and anp non-LOCA non-LOCA safety analyses are performed safety analyses performed to justify that the nuclear justify that nuclear power plants can be safely shut power plants down following shut down following postulated accidents.

postulated accidents. Although Although these analyses do these analyses specifically consider do not specifically consider failure (i.e.,

(i.e.,. complete complete severence) of aa CRDM CRDMnozzle, consider events that they consider nozzle, they have more limiting consequences.

consequences. LOCA analyses analyses typically breaks In postulate breaks typically postulate in RCS pipes from from those those within the plant makeupmakeup capacity up to and including including aa double-ended guillotine double-ended guillotine break break of the hot leg feg to demonstrate acceptable core demonstrate acceptable core coaling in cooling in the the short short term term asas well as the the long Non..;LOCAsafetyanalyses long term. Non-LOCA safety analyses postulate a control specifically postulate specifically control rod ejection accident. although the ejection accident, CRDM nozzle the CRDM nozzle remains intact. The ejection event The rod ejection event postulates postulates that the CRDM flange bolts fail that the and the control and control rod is ejected ejected outout of of the CRDM CRDM housing.

housing. These These plant safety analyses are analyses are reviewed reviewed in the following paragraphs to determine following paragraphs determine if a more more substantial safety substantial safety issue exists based on the leaks that have been observed at have been ONS-1, ONS-2, ONS-3, and ANO-1.

ONS-1, AN0-1. Where applicable, additional margin is applicable, additional identified to furthersupport further support plant restart restart and continued safe safe operation.

described in the previous sections, once a crack initiates, it is estimated As described estimated that it may take up to six years for it to migrate through through the CRDM components and components and of such Detection *of such minor undetectable rates. Detection begin to leak at undetectable minor leaks that grow at slow rates is is by visual inspections of the CRDM nozzles noted with the ONS-nozzles as noted ONS-ONS-3, and ANO-1 1, ONS-2, ONS-3,and ANQ-1 outages. These inspections of the These routine inspections the potentially affected areasareas will identify before the weld initiated well before identify if any leak has initiated weld component could fail catastrophically. The detected or component detected cracks have have grown predominantly predominantly in the axial direction, although some circumferential direction, altho6ugh circumferential cracks have have observed near the weld. These as-found been observed circumferential and axial cracks as-found circumferential evaluated, and it was concluded that the structural have been .evaluated,and integrity of the structural integrity component retains sufficient margin to ensure continued safe operation sufficient margin operation of the plant. In addition, the maximum projected growth rate from the boric ~cid acid penetration from a minor corrosion of the RV head penetration corrosion minor leak would not propagate into adjacent CRDM nozzle nozzle failures. Therefore, simultaneous catastrophic failure of simultaneous catastrophic multiple nozzles will not be postulated.

Since failure .of CRDM nozzles multipleCRDM of multiple considered credible, the primary nozzles is not considered primary concern is the failure of a single nozzle. This unlikely, yet postulated concem postulated failure leads to RCS inventory loss and less core shutdown margin margin for the plant safety analyses. These aspects are addressed in the following paragraphs paragraphs relative to the consequences already Included In In the existing LOCA and non-LOCA plant plant safety analyses.

Loss-of-CoolantAccident Loss-at-Coolant Plant LOCA analyses do not specifically analyze the potential failure of the the reactor vessel or any of the attachedattached nozzles, but they do postulate break sizes from 0.01 ff ft 2 to 14.2 ft in area in any RCS pipe. A 14.2 *trin break in a CRDM from a A break crack that formed, propagated catastrophically propagated without detection, and failed catastrophically Page 19 of 56 NRC004-1190 NRC004-1190

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RV Head Nozzle and Weld Safety Safety Assessment Assessment 51-5012567-01 51-5012567-01 would be bounded bounded by the ReS RCS inventory losseslosses considered in the existing plant LOCA LQCA analyses. Also, this break break location location is favorable from a core cooling cooling standpoint, in that it is on the hot side of the core, such that no emergency standpoint,inthat emergency core cooling system (ECCS) fluid is bypassed directly out of the break. That means

s bypassed means the ECMS that all .the ECCS liquid isIs available for core cooling.

COOling. The core shutdown for this this is assured by the insertion event isas.sured insertion of the remaining remaining control rods.

rods, augmented augmented by the soluble boron reactivity control via the boron in the ECCS injection fluid. fluid..

Despite the fact that the existing LOCA analyses bound the CRDM nozzle failure LOCAanalyses with respect to inventory inventory loss, there there remains additional margin based on the the

.credited rod worth and the RCS leakage detection detection systems. In the small break LOCA analyses, minimumminimum control control rod worths are credited. The control rod of assumed to be stuck out of the core, and only a fraction of the highest worth is assumed the remaining worth is used in demonstrating demonstrating that at least a 1 percent percent shutdown margin exists at hot zero power power conditions.

The RCSleakage RCS leakage detection detection systems are required required by the plant technical specifications to detect specifications unidentified leak rates of 1 gpm or greater. IfIf the leak detect unidentified rate is higher, the plant wilJbe will be shut down, and a controlled cooldown will be controlled cooJdown be initiated. The makeup system will provide sufficient inventory and boron control.

Insertion will Insertion of the control rods will not be inhibited, inhibited, and the core reactivity will be core reactivity be controlled. Following reactor shutdown, the consequences controlled. FollOwing consequences of a CRDM nozzle nozzle failure are decreased, decreased, thereby providing additional assurance assurance that a safe safe shutdown shutdown is not compromised by the leakage leakage that has been found or postulated postulated to propagate during during a single operating cycle with a leak in inaa CRDM CRDM nozzle.

Non-LOCA Safety Analyses Non-LOGA The plant plant non-LOCAsafety non-LOCA safety analyses, for which consequencesconsequences can be more severe ifif the core is not completely completely shut down, assume that the highest worth worth control rod is stuck out of the core, and at least a 1 percent .shutdown shutdown margin margin exists at hot zero power conditions. Also, the consequences consequences of a control rod accident (CREA) are explicitly ejection accident expliciUy analyzed analyzed and included in the individual Final Safety Analysis Report plant Rnal Report (FSAR). Limitations are also imposed on each core design to limit the worth of any ejected control rod worth at hot full power to much less than the value assumed in the accident a value much accident analyses.

The.s~andard The standard NRC-approved NRC-approved methodology methodology (for Framatome Framatome ANP) consists of (1 (1))

calculating calculating the maximum maximum single ejected ejected rod worth throughout throughout cycle life, (2) (2) verifying that the limits bound these maximummaximum worths after augmenting by a 15 percent uncertainty, uncertainty. and (3) verifying that the core operating operating (rod index) limits limits preserve the calculational calculational* basis of the maximum worth. Because the typical analysis methodology methodology uses the core average power response. response, the results of the the calculation are sensitive to the total amount of reactivity inserted, not the number Page Page 20 20 of of 56 56 NRC004-1191

• • NON-PROPRIETARY NON-PROPRIETARY **** **** 51-5012567-01 51-5012567-01 RV Head Nozzle Nozzle and Weld Safety Assessment Assessment of control rods ejected. Consequently, Consequently, the existing analysis will rernainbounding remain bounding for any number of ejected ejected control rods, provided provided the total reactivity reactivity inserted into the core remains less than the values analyzed and reported in the FSAR. This This provides provides additional additional margin, margin, such that the consequences consequences for the unlikely unlikely failure of singleCRDM a single nozzle will not be more severe .than that already considered CRDM nozzle considered by each new fuel cycle for a limiting limiting control ejection accident control rod ejection accident scenario.

scenario. .

9.0 Risk Assessment for CRDM Nozzle Cracks Cracks The purpose purpose of this section section is to provide a risk analysis analysis to supplement and support the deterministic deterministic safety assessment.

assessment. The other sections of this safety assessment report describe the traditional traditional engineering assessment assessment of the CRDM CRDM nozzle cracks, including deterministic deterministic issues such as the impact upon safety margins and defense-in-depth.

defense-in-depth. This determ!nistic analysis provides the source deterministic analYSis material upon which the risk assessment is based. This risk analysis estimates estimates the core damage damage frequency frequency (CDF)

(COF) associated operation with potentially associated with operation potentially undetected CRDM nozzle undetected recently at ONS.

nozzle cracks, such as those found recentfyat 9.1 Potential Risks from CROMCRDM Nozzle Nozzle Cracking assodated with the p~s;bility Potential risks associated possibility of undiscovered undiscovered CADMCRDM nozzle nozzle cracks cracks include:

• • LOCA due to CRDM nozzle detachment nozzle rupture or detachment

• • Anticipated Transient Without Scram AntiCipated Scram (ATWS) due to rod ejection accident accident (one or multiple)

• " A ATWS TWS due to loose parts blocking control rods rods

• " Damage Damage due to CRDM and nozzle missile missile during accident accident LOCA is Of these, LOCA considered to be the most important from a core damage is considered damage perspective.

frequency or risk perspective.

The random nature nature of crack initiation and growthgrowth makes it highly highly unlikely that multiple circumferential circumferential cracks will reach critical size in different CRDM nozzles nozzles at the same time. This assertion is made because because the mean-time-to-failure mean-time-to-failure in given CRDM nozzle population is randomly (and widely) distributed. Even if any givenCRDM there were several CRDM CRDM nozzles with unrevealed unrevealed degradation, degradation, the loadsloads administered administered during a plant transient transient would not impact them in identical ways.

Because of nonhomogeneous Because nonhomogeneous crack initiation and growth, growth, one CRDM nozzlenozzle failure time would precede the other(s). The recent B&WOG experience B&WOG plant experience Page 21 of 56 56 NRC004-1192 NRC004-1192

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• • • • NON-PROPRIETARY -** 51-5012557-01 51-5012567-01 RV Head Nozzle Nozzle and Weld Safety Assessment Assessment (see Sections 2.2 and 9.2) supports supports this assertion. The evidence evidence of crack crack extent for the observed 00 fortheobsetved OD circumferential circumferential cracks above the J-groove J-groove weld indicates a random distribution distribution of crack lengths. Therefore, simultaneouscrack.:lnitlated simultaneous crack-initiated failures of redundant CRDM nozzles are very unlikely, and the risk from multiple redundantCRDMnozzies multiple CRDM nozzle failures due to cracking is very small.

CRCM nozzle In addition, the plant shutdown shutdown margins (if evaluated realistically) are such that several several CRDM nozzlenozzle failures could be tolerated before before the risk would increase increase over that otaof a single CROMnozzie CRDM nozzle *failure.Even failure. Even with conservative conservative success success criteria for reactor trip, trip. two or three CRDM nozzle failures could easily be be tolerated from aa reactivity reactivity standpoint-standpoint Therefore, it Is is concluded concluded that the the reactivity accidents reactivity ejection or control rod blockage by loose accidents (rod ejection loose parts) are not credible .risk risk contributors, because because of the number of simultaneous CRDM CROM nozzle nozzle failures that would be required.

failures required.

Missiles generated generated by CRDM CROM nozzle failures are also not credible credible risk contributors. Even in the unlikely event of a CADM CRDM nozzle detachment.detachment, the the missile shields will prevent consequential consequential damage to the reactor building building or other safety systems.

Therefore, impact that will be addressed and quantified is the risk from a Therefore. the risk impact LOCA. This analysis will estimate incremental CDF due to a LOCA estimate the incremental LOCA caused by a CRDM nozzle nozzle that may fail during operation due to undiscovered undiscovered cracks.

9.2 Identification Identification of CADM CRDM Nozzle Nozzle Cracks that arearea a Risk Concem Concern Of particular particular concem concern are circumferential circumferential cracks above the weld. A circumferential circumferential crack sufficient extent may cause a large crack of sufficient large leak or a LOCA LOCA duedue to gross structural failure (net-section collapse) of the CAOM CRDM nozzle pressure pressure boundary.

boundary. The 00 OD of the CRDM nozzle nozzle just above the J-groove J-groove weld (which is normally dry) is the only region on the CRDM nozzle pressure boundary where where there is high axial stress relative to hoop stress. This region is susceptible to to circumferential circumferential PWSCC cracks only if there is a source source of primary water to the nozzle penetration annulus, such as might occur nozzle penetration occur it throug~wall (TW) if there is a through-wall axial axial, crack initiated from the 10 ID of the CRDM CRDM nozzle or a crack in the J-groove J-groove weld.

Although axial CRDM nozzle cracks and J-groove weld cracks have occurred, they are not likely to result in a significant LOCA directly. The primary risk concern with the axial crackscracks and weld cracks is that the primary water leakage leakage through the crackcrack can provide moisture to the CADM CRDM nozzle exterior and promote 00 OD PWSCC.

PWSCC.

Page 22 of 56 56

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... NON-PROPRIETARY **** **** 51-5012567-01 51-5012567-01 RV Head Nozzle and Weld Safety Assessment A$sessment Recent Recent B&WOG B&WOG experience experience (see Section 2.2) has has included included several axial crackscracks propagating propagating through the J-groove J-groove weld or the area of the CROM CRDM nozzle nozzle near near the the weld. At ONS- 3, there were nine CRDM nozzles nozzles with TW axial cracks above cracks above the the J-groove welds j-groove W$lds or in the welds themselves. These cracks provided provided a' a path for primary water to the 00 primary OD of the CRDM nozzle (in the annulus annulus area just above the subsequent inspection weld) where subsequent inspection .indicated indicated that three of these had indications indications OD circumferential of 00 circumferential cracks above the weld. At ANO-1, there was an axial 00 AtANO-1, OD crack in the nozzle nozzle below the weld (i.e., in the area that is normally wetted) that extended to above the weld on the nozzle OD, extended 00, thus wetting the 00 OD area above above the weld. At ONS-1, ONS-1, there was a crack through the J-groove JiVoove weld of a CRDM CROM nozzle that wetted the OD 00 of the nozzle in the annular region above the weld. weld.

And at ONS-2, four CRDM nozzles nozzles werewere found with axial cracks that caused leakage to the annular region. Of these,one leakage these, one had indications indications of an 00 OD circumferential crack above the weld. This experience circumferential experience suggests that there is a risk of a LOCA-sized LOCA-sized CROMCRDM nozzlenozzle failure from 00 OD circumferential circumferential cracks that may be initiated due to primary may primary water leaking into the annulus area from undetected ID-initiated TW axial, other OD-initiated undetectedlo-initiated OD-initiated axial,or axial, or J-groove J-groove weld weld c

cracks.racks.'

The four above-the-weld above-the-weld 00 OD circumferential circumferential cracks at ONS-2 ON8-2 and ONS-3 were repalred repaired along with the other crack indications. Excavations Excavations to clear clear these these indications extended extended up to 180 180 degrees circumferential direction, degrees in the circumferential direction, and complete complete characterization characterization of the indiCations indications was not recorded due to the the aggressive aggressive nature nature of the excavations.

excavations. However, subsequent examinations of subsequent examinations ultrasonic test (UT) data taken before before the excavations have

the excavations have indicated indicated thethe circumferential circumferential extents of the cracks to be approximately approximately 36 degrees degrees (ON8-2 (ONS-2 18), 66 degrees (ONS-3 nozzle 23), 59 degrees nozzle 18). degrees (ONS-3 nozzlenozzle 50), and degrees (ONS-3 nozzle 165 degrees nozzle 56) (see Section 2.2).

It is also possible that an ID4nitiated 10-initiated crack circumferentially to fail the crack could grow circumferentially the CRDM nozzle pressure CROM boundary directly. These cracks have been considered pressure boundary considered in this risk assessment. However, the operating history and probabifistic probabilistic fracture fracture mechanics analysis of ID-initiated mechanics circumferential cracks indicates IO-initiated circumferential indicates that the the likelihood of this failure mode mode is very small due to the nature of the stresses on the ID CRDM nozzle. To support this assertion, 10 of the CROM assertion,. a probabilistic probabilistic fracture mechanics prediction prediction was made (Reference(Reference 29) of thelD-initiated the ID-initiated TW 1W crack frequency frequenCy using the CHECWORKS CHECWORKS computer code (Reference (Reference 30), the EPRI tool for predicting predicting time to Alloy 600 PWSCC. The CHECWORKS CHECWORKS analysis shows that the expected frequency expected frequency of ID-initiated circumferential cracks is much less than 10-initiated circumferential the expected expected frequency of IO-initiated ID-initiated axial cracks. For the worst-case worst-case B&WOG B&WOG plant, CHECWORKS CHECWORKS predicts predicts a median cumulative cumulative probability of 0.07 over the the (60 year) plant life of getting an ID-initlated ID-initiated above-the-weld TW circumferential above-tha-weld TWcircumferential crack. This is a frequency frequency of approximately 0.002 per reactor-yearifreactor-year if averaged averaged over the remaining plant life. That frequency frequency is insignificant insignificant relative to the the probability of axial cracks probability cracks that may contribute contribute to OD 00 PWSCC (which is discussed Page 23 of 56 NRC004-1194 NRC004-1194

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• • • • NON-PROPRIET NON-PROPRIETARY**** MY **- 51-5012567-01 51-5012567-01 RV Head Nozzle and Weld Safety Assessment Assessment further in Section Section 9.3.1). Therefore, the focus of the risk assessment assessment is on the the scenarios scenarios for circumferential circumferential 00 OD CRDM CROM nozzle cracking, and the risk estimated estimated for OD-initiated aD-initiated circumferential circumferential cracks is considered considered representative representative of 01 the overall '

risk.

With respect respect to impact impact upon risk, the only CROM CRDM nozzle cracks that are risk are. those where detectable significant are, detectable symptoms of the degradation degradation are not identified identified (and acted upon) prior to total failure of the nozzle. The risk analysis analysis discussed below discussed below estimates estimates the probability that CRDM CRCM nozzle failure will occur occur before a successful vi,sual visual inspection detects detects telltale boron crystals crystals on the the exterior of the reactor vessel head. '

9.3 OD OD Circumferential Circumferential Crack Risk The risk analysis focuses focuses on scenarios scenarios in which an 00 OD circumferential circumferential crack can can grow to failure, causing a LOCA. The OD circumferential crack grows on the 00 circumferential the CROM CRDM nozzle pressure boundary as a result of PWSCC caused by CRDM pressure boundary CROM nozzle of J-groove J-groove weld leakage leakage that wets the exterior of the CAOM CRDM nozzle nozzle in the the annulus around the head penetration.

head penetration. The incremental incremental core damage frequency damage for a LOCA induced induced by 00 OD circumferential circumferential CRDM nozzle cracking cracking Is is the product product of the following factors:

• - Frequency Frequency of weld or nozzle nozzle leak that wets 00 OD of CRDM nozzle nozzle in the the susceptible location

• • Probability that CRDM nozzle nozzle leakage is undetected leakage Is undetected

• Time-dependent Time-dependent probability prQbabilitythatthat total failure of CRDM nozzle will occur fai/ureof occur due to undetected undetected crack initiation initiation and growth on nozzle nozzle 00OD

• " Probability of core damage damage from resulting LOCA LOCA These events are shown as headers headers on the event tree (Figure 5) in In which sequences that result in core damage sequences damage are shown. Estimates*

Estimates of the event tree probabilities and Initiating initiating event frequency provided in the following frequency are provided following sections.

Probability of Weld or Nozzle 9.3.1 Probabilitv Nozzle Leak In this section, the frequency of CRDM nozzle leaks that may wet the 00 OD above above the weld is estimated. It isassur:ned is assumed that some CROM CRDM nozzles may be in service service with near-lW near-TW nozzle nozzle cracks or weld cracks that may surface in the next fuel cycle. The CRDM nozzle cracks of interest are those above weld axial cracks cracks Page 24 of 56

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NON-PROPRIETARY 51-5012567-01 Nozzle and Weld Safety Assessment RV Head Nozzle and weld cracks that may leak primary primary water to the exterior exterior of the CRDM nozzle nozzle in the annulus region of the RV head penetration.

The CADM CRDM nozzle nozzle leak rate has been estimated from the recent recent' inspection inspection experience at ONS-1, experience ONS-1, ONS-2, ONS-a, ONS-3, and ANO-1.

ANO-1. At these fout four plants, boron crystal deposits indicated indicated leakage leakage at 15 CRDM nozzles,nozzles. including including one at ONS-1,ONS-1, four at ONS-2, nine at ONS-3, ONS-a, and one atANO-1 at ANO-1 (see Section 2.2). Itis It is uncertain how how long these CRDMCAOM nozzles nozzles have been leaking. For the purpose of estimating a leak frequency, it will be assumed* assumed that half half of these 15 leaks appeared during the most recent fuel cycle and half in the previous appeared previous fuel cycle.

cycle. ItIt is likely that some of these leaks actually present in leaks were actuaJly earlier refueling In earlier refueling outages, but were not identified as nozzle leaks at that time. Therefore, 15 15 leaking CRDM nozzles nozzles in approximately approximately twelve plant-years plant-years (four plants plants times two cycles cycles times 1.5 years per fuel cycle), gives an average frequency of average frequency approximately approximately 1.25 leaking leaking CRDM nozzles per reactor-year.

A prediction was also made (Reference (Reference 29) of the lD-initiatedTW ID-initiated TW crack frequency using CHECWOAKS frequency CHECWORKS (Aeference (Reference 30). CHECWORKS is an empirical 30). CHECWORKS code, and the recent inspection were taken Into inspection results were into account account by adjusting the the crack reference times. The results *of crack initiation reference of the CHECWORKS CHECWORKS analysis of 10- ID-initiated initiated cracking are dominated by the contribution contribution from axial cracking, which was over two orders of magnitude magnitude more likely than circumferential circumferential 10 ID cracking to cause 1W crack above the weld. The cause a TWcrack "resultsof this analysis predict a median The'tesults frequency otlO-initiated frequency of ID-initiated TW cracks the J-groove weld of 0.52 per reactor-cracks above theJ-groove averaged over the remaining plant life (assuming 60 year life).

year averaged life). However, the CHECWORKS CHECWOAKS analysiS analysis is Is for ID-initiatec..

ID-initiatec_ nozzle crackingcracking only.

CHECWORKS CHECWORKS (as it is now configured) is not designed designed for OD-initiated cracks above above the weld or for J-groove J-groove weld cracking. Hence, Hence. as a prediction for CADM CRDM nozzle leak frequency, it may underestimate. Therefore, Therefore, to be conservative conservative the the mean mean value of 1.25 CRDM nozzle CRDMnozzleleaks leaks per reactor-year, which which was estimated estimated from the plant experience, experience, will be usedused in the risk assessment.

9.3.2 Probability that CRDM Nozzle Leakage Leakage is Undetected Undetected A human reliability analysis analysis (Reference 31) has been performed to estimate the (Reference 31) the human error probability (HEP) for the utility's inspection personnel personnel failing to detect boron crystal deposits deposits on the RV head that are indicative indicative of a CRDM nozzle leak. CRDM nozzle nozzle leakage will be detectable detectable through the accumulation accumulation of boron crystals on the top of the RV head around the base of the atfepted affected CRDM nozzles. It is assumed that any CROM CRDMnozzles. CRDM nozzle crack crack is undetectable undetectable until a through-wall through-wall crack crack (or weld crack) deposits boron crystalscrystals on the exterior of the the RV head.

RVhead.

~

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Nozzle and Weld Sa~ty Safety Assessment 51-5012567-01 For OD-initiated above-the-weld For above-the-weld cracks, the the fracture fracture mechanics model that is human reliability analysis is how long most relevant to the hurnanrellabillty long itit takes, once wetted, an 00 for an OD crack to initiate and and grow to the critical size size for CRDM nozzle nozzle failure.

time (estimated in Section This time. Section 9.3.4) will Indicate indicate how many opportunities opportunities (refueling outages) there may be to detect the boron crystaJsbefore crystals before total failure of the CROM CRDM nozzle. Another factor important to to the risk assessment assessment is when the boron crystal deposition will be visible relative to the growth of the the circumferential cracking.

circumferential The 00OD PWSCC failure mechanism mechanism requiresrequires aa moist environment from the the presence of primary water in in either the liquid liquid or steam state. Primary Primary water fromfrom aa leaking CRDM nozzle nozzle or weld will be deposited into the nozzle penetration penetration annulus. During steady state operation there isa is a small radial clearance in the the annulus above the weld to the surface of the RV head (see (see Appendix A). A). The primary leakage into the annulus may initially be very small, as might be the case for a pinhole leak, or somewhat larger, but there can only be PWSCC when there leakage to keep the annulus area moist. Very small leaks is a sufficient rate of leakage appropriate environment are not likely to provide the appropriate environment in the annulus initially, initially, temperatures and pressures on top of the reactor vessel. The considering the temperatures The rate of boron crystal deposition will also be dependent dependent upon the size of the leak.

Moderate-sized leaks will generate Moderate-sized generate boron crystals rapidly. For smaller leaks, there may be some time before significant boron crystals accumulate. However, However, as the boron crystals build up in and aroUnd aro~nd the annulus, their presence presence will tend tend to trap moisture below. It is also possible that for a small leak there there may be intermittent intermittent "leak plugging" and a weeping weeping type leak (see Section Section 5) as the the buildup buildup of boron crystals intermittently "vents." Hence, it is reasonable reasonable to environment required for the initiation of PWSCC conclude that the environment PWSCC on the 00 OD of the CRDM the CRDM nozzle nozzle (i.e., above the weld), whether it be from a small or moderate (i.e., above moderate

leak, leak, will coincide roughly will coincide roughly with with the the presence presence of of visible boron crystal visible boron crystal deposits.

deposits.

Reactor Vessel Head Inspections 9.3.2.1 Reactor Inspections As a result of Generic Generic Letter 97-01 (Reference (Reference 4), the B&WOG B&WOG licensees have licensees have made a commitment commitment to perform perform timely inspections inspections of CRDM nozzles (and other CRDM nozzles vessel closure head penetrations).

penetrations). This commitment commitment is maintained maintained by permanent permanent addition addition of a task item/work item/work order into into the refueling refueling outage outage schedule program.

Discovery Discovery of (new) boron on the head head will result in the finding being placed placed in the licensee's Corrective licensee's Corrective Action Action Program (CAP).

Program (CAP).

CRDM nozzle flange leaks CROMnozzle leaks have have occurred on several past occasionsoccasions at B&WOGe&WOG plants. Boric acid crystal Boric add crystal buildup buildup from thesethese leaks may have masked masked indications indications of CRDM CRDM nozzle nozzle leakage leakage in the past, and may have have contributed contributed to the the exterior exterior circumferential circumferential OD cracks at ONS :not being detected cracks ONSnot being detected by an an inspection inspection sooner.

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• • • • NON-PROPRIETARY - ** - 51-5012567-01 RV Head Nozzle and Weld Safety Assessment Assessment CRDM nozzle CRDM nozzle flange leakage in the past was not considered considered to be unusual; however, once discovered, the CRDM nozzle flange was repaired to stop the CRDM nozzle the leakage.

leakage. Part of the repair processprocess was to replace the gasket. The B&WOG B&WOG licensees licensees have been gradually gradually repairing flanges and replacing gaskets gaskets since about May 1989.

1989. To date, nearly nearly all of the B&WOG plantCRDM plant CRDM nozzle flange gaskets have have been replaced replaced with a stainless stainless steel/graphite steel/graphite gasket, which, according to operating experience, experience, are less prone to leakage. leakage. The number of CRDM nozzle flanges that still have the old gaskets is currently gaskets ;s currently quite small (total (total of about a dozen over allot all of the B&WOG B&WOG plants). Any flange leakage leakage from one of these few remaining remaining old-style gaskets gaskets would be quite evident, and would be promptly addressed.

Over the last five to seven years, the RV head inspections inspections have have become become increasingly meaningful because increasingly more meaningful because of utility efforts to clean the head of boron deposits deposits resulting from past CRDM nozzle flange leakage leakage and other sources. A clean clean RV head will make new boron crystals crystals at the nozzle penetrations more nozzle penetrations more evident, and reduce the likelihood leakage will be missed or masked by likelihood that the leakage other sources sources of boron on the RV head.

The method method of RV head inspection inspection for indications of boron varies among the B&WOG plants. The methodsmethods vary from a simple simple visual inspection inspection to the use of mobile AV a mobile RV head -robot robot with an attached attached video camera. However, none none of the the B&WOG plants insulation djrectl~

plants have inSUlation directly on the reactor vessel head that may impact visual inspections. With all of the methods, the RV head inspection visual Inspections. inspection process straightforward, such that a written procedure process is simple and straightforward, procedure is not necessary for a successful necessary successful inspection. For the visual method, method, the RV headhead is observed through eight observed acces~ panels in the service structure eight or nine access structure with a high inspector would be from aCRDM intensity portable light. The farthest an inspector a CRDM nozzle nozzle completeness, the inspection is five feet. To ensure completeness, inspection is carried carried out with a paper map of CRDM nozzle nozzle locations. The visual inspection inspection method requires requires approximately two hours to complete. Other methods, such as use of a approximately boroscope (i.e., camera on a stick) or RV head robot, result in aa permanent permanent inspection on videotape. These record of the inspection These methods also rely on the use of a paper map of CRDM nozzle locationslocations to ensure completeness of the inspection.

9.3.2.2 Estimate Estimate of Human Human Error Probability Probability for Visual Inspections Inspections HEPs have been HEPshave been estimated for failure of the visual inspections using a combination of the Human combination Human Cognitive Reliability Reliability Model (Reference (Reference 32) and Swain Guttman's Handbook and Guttman's Handbook (Reference (Reference 33). Since, visual inspections will occur occur with each refueling refueling outage.

outage, aa time-dependent time-dependent failure probability is estimated estimated inspections to occur at two-year intervals. This is conservative considering the inspections conservative refueling cycles range between 18 and 24 months.

since refueling Page 27 of 56 NRC004-1198 NRC004-1198

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• -*NON-PROPRIETARY - *- 51-5012567-01 RV Head Nozzle Nozzle and Weld Safety Safety Assessment The human reliability analysis considered considered three ways in which the inspection inspection process can fail to detect the boron boron crystals that are indicative indicative of aa CRDM nozzle nozzle leak. These These include failure to conduct conduct the inspection, failurefaiJureto observe the boron to observe crystals on the RV taead head when present, or failure to identify boron crystals crystals resulting from aa CROM CRDM nozzle leak due to masking masking by other sourceS sources of boron (i.e., from CRDM nozzle flange leakage). leakage).

The human reliability analysis estimates that the HEP for failure of the visual inspection to detect signs of CRDM nozzle inspection nozzle leakage opportunity is leakage at the first opportunity 2

6.0x1 0-. The humar1 6.0x10-

• analysis also estimates the failure probability human reliability analysis probability for a second and third inspection (spaced at refueling outage intervals) second Intervals) of finding the the same same leaking CRDM nozzle, given failure of the previous inspection(s). inspection(s). The The probability of repeatedly failing to detect the boron deposits deposits at consecutive consecutive inspections inspections (assumed two-year intervals)isa intervals) is a dependent relationship.

relationship. The The impa~t impact of this dependency dependency (aside from the crack growing larger) is that the boron boron deposit deposit will be more prominent prominent and more difficultdifficult to miss at the next inspection.

inspection.

However, there is also the possibility that errors made in In previous inspections will be repeated, repeated, the most important Important error being failure to perform inspection.

perform the inspection.

The human reliability analysis analysis balances these competing dependencies, dependencies, and appropriately appropriately adjusts the HEP with each subsequent subsequent outage. After failure of the the first visual inspection, the dependency repeatedly failing to perform dependency for repeatedly perform the the inspection inspection is conservatively conservatively assumed assumed to be slronger stronger than the dependency dependency of the the boron deposit being more more evident with lime, !ime, thus causing causing the HEPHEP to increase increase with additional opportunities. The human reliability analysis additional opportunities. analysis estimates estimates that the

- HEP for failure to detect the CRDM nozzle leakage leakage on the second opportunity opportunity is 6.5x1 6.5x10- 0-2 and that It it is 0.11 for the third and each subsequent outage. These each subsequent These HEPs ,are are conservative considering the increased c"'1servativeconsidering increased future emphasis on effective effective visual inspections inspections of the reactor vessel head penetrations.penetrations. Conservative Conservative HEPs HEPs have been used to encompass encompass the uncertainty that is generally generally present present in HEP HEP estimates.

9.3.3 Probability Probability of OD Initiation 00 Crack Initiation The time-to-OD-crack-initiation, time-to-OD-crack-initiation, once the exterior of the nozzle is wetted with with primary primary water, Is unknown.

unknown. Computer codes used to predict time-to-PWSCC initiation are unreliable unreliable for 00 OD PWSCC becausebecause the environment environment on the exterior exterior of the CRCMCRDM nozzle (i.e., exterior to the pressure boundary) may be different nozzle (I.e., different than on the 10 ID of the nozzle, especially especially in terms of boron concentration concentration and and length length ,of of time wetted. Therefore, to be conservative, the risk analysis assumes assumes that the time-to-OD-crack-initiation time-to-OD-crack-initiation is zero for all CRDM nozzles nozzles with exterior exterior primary water wetting.

primary '

This approach approach is conservative conservative with respect to the-observations the observations of the 15 leaking leaking CRDM nozzles nozzles that were recently found at ONS and ANO-1. Only four of these were recently ONS ANQ-1.

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.RV Head "Nozzle RV Head Nozzle and Weld Safety Safety Assessment Assessment nozzles had CRDM nozzles indications of OD had indications 00 circumferential cracking above the weld.

circumferential cracking weld. It unknown specifically is unknown specifically how longlong each of CROM nozzles of these CRDM nozzles has has been been leaking leaking or whether whether OD 00 cracks cracks would have have initiated others if initiated on the others leakage had continued jf leakage continued undetected. Therefore, undetected. since aa valid Therefore, since time-dependent model valid time-dependent model for OD 00 PWSCC PWSCC crack Initiation is crack Initiation Is unavailable, conservative to assume unavailable, it is conservative assume that OD 00 crack crack initiation initiation will occur occur in 1100%00% of the CRDM nozzles that have CROM nozzles leakage into the have leakage annular the .annular region above the weld.

approach (100%

This approach crack initiation with (1 00% crack time~to-Initiation) bounds with zero time-to-Initiation) bounds the the uncertainty associated uncertainty associated with the lack lack of of probabilistic fracture mechanics probabilistic fracture mechanics data for for 00 crack initiation.

PWSCCcrackinitiation.

OD PWSCC 9.3.4 Time to Total 9.3.4 Total Failure CROM Nozzle failure of CRDM Nozzle A probabilistic mechanics analysis (Reference probabilistic fracture mechanics performed to (Reference 34) was performed determine the probability determine net-section failure of CRDM probability of net-section nozzles after initiation of CROMnozzies above-the-weld OD above-the-weld 00 circumferential circumferential cracking.

cracking. The probabilistic probabilistic fracture fracture mechanics model model was built around the deterministic deterministic crack growth growth model described in Section 3.3.2. The crack growth (Reference 18) described (Reference growth model uses uses thethe model with worst case stresses.

Peter Scott model probability of gross net-section stresses. The probability net-section determined by performing failure is determined Monte Carlo simulation on a typical performing a Monte B&W-typicalB&W-designed CRDM nozzle by varying designed parameters of crack growth and varying the defining parameters size used in .the the deterministic fracture mechanics analysis. Available industry data were were used to define distributions for key variables define distributions conservatism was variables and conservatism was where the data were sparse.

used where For the initial flaw distribution, the calculation performed using parameters calculation was performed parameters representative of the nozzle representative nozzle cracks found at ONS. For example, example, UT exams of above-the-weld 00 the four above-tha-weld circumferential crack OD circumferential crack indications ONS-2 (nozzle indications at ONS-2(nozzle

18) and ONS-3 ONS-3 (nozzles 23, 50, 56) .indicate circumferential extents of indicate circumferential approximately 36 approximately degrees, 66 degrees, 59 degrees, 36* degrees, degrees, and 165 degrees, degrees, respectively {see 2.2). It is unknown whether each of these cracks grew (see Section 2.2}.

from a single 00 site, or from several initiation sites that linked together OD initiation site,or circumferential 00 to form a long circumferential Therefo~e, the initial flaw size OD surface crack. Therefore,

~

used in the Monte Carlo simulation is a shallow semi-elliptical semi-elliptical flaw with a circumferential circumferential extent uniformly distributeddistributed between zero and 180 degrees.

Postulating a single flaw with an initially long circumferential extent is an circumferential extent approximation of the _possibility approximation possibility of multiple linked initiation sites. The ONS plant plant experience is consistent with a uniform distribution of initial flaw extent ext~t and this this approach is reasonable in light of the sparse industry data available for 00 OD flaw distributions. A practical upper limit for this initial flaw distribution is a circumferential extent of 180 circumferential 180 degrees, which is related to the nature of the stresses on the surface of the CROM CRDM nozzle above the weld. On the nozzle 00 OD above the weld, crack initiation in the circumferential direction may be driven by by Page 29 of 56 Page NRC004-1200 NRC004-1200

RV Head Nozzle

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Nozzle and Weld Safety Assessment Assessment 51-5012567-01 51-5012567-01 the axial bending bending stresses that are related to the fit zones, and these are different thl:! proximity of the weld and shrink different on the uphill and downhill downhill side of the nozzle.

E This approach approach of postulating an initial flaw as long as 180 180 degrees in circumferential circumferential extent bounds the uncertainty uncertainty from scarcity probabilistic scarcity of probabilistic mechanics data for OD fracture mechanics 00 flaw distributions initiation sites.

distributions and multiple .initiation Another source of uncertainty uncertainty.isis the crack growth rate for OD-initiated PWSCC.

OD-initiatedPWSCC.

The flaw growth rate distribution growth rate distribution used in the Monte Monte Carlo simulation simulation ýisis based upon industry data for PWSCC. ParametersParameters affecting affl:!cting growth growth rate, such as stress intensity and temperature, were distributed Intensity distributed in the Monte Monte Carlo model model to address address uncertainty. Figure 6 illustrates illustrates the resulting crack growth rate distribution that was assumed in the Monte Carlo carlo simulation. However, it is Is unknown whether the the difference in environment environment between the nozzle exterior exterior and interior rnay affect the may affect the growth growth rate for 00 OD PWSCC. The approach approach used in this risk assessment assessment to ensure uncertainty associated with crack ensure that the uncertainty crack growth rate is bounded, is to benchmark the Monte Monte Carlo simulation results for time-to-TW time-to-lW crack against the the plant observations. If If the crack growth growth data are reasonable, the Monte-Carlo Monte-Carlo simulation predict TW crack times consistent simulation should predict consistent with the plant observations.

observations.

The results of the Monte simulation for TW cracking are illustrated Monte Carlo sImulation illustrated by the the histogram histogram shown in Figure 7. The Monte Monte Carlo simulation simulation results are consistent consistent with the plant experience. Of the 15 leaking CROM CRDM nozzles found at ONS and ANO-1, two had above-the-weld ANO-1, OD,,Pircumferential cracks above-the-weld OD,pircumferentiai cracks that were TW or almost TW (ONS-3 (ON5-3 nozzles nozzles 50 and 56). To reach reach the equivalent percentage of equivalent percentage TW cracks In in the Monte Monte Carlo simulation (i.e., 13.3% of the samples) required simulation (Le.,

4.2 years. Anecdotal evidence suggests that some some of the nozzles at ONS may have been leaking for as much as 5 to 10 years. Therefore, the results of the required the I

Monte Carlo Carto simulation appear to be reasonable reasonable or conservative cons.ervative with respect respect to experience.

The Monte Monte Carlo Carlo simulation simulation was used used to grow the initial flaws to failure using the the crack growth model described described in In Section 3.3.2 and stress distributions distributions that are characteristic of the most exterior characteristic exterior nozzles highest angle nozzles (i.e., the highest angle of penetration).

For this analysis, failure is Is defined defined as insufficient insufficient ligament to meet ASME Code primary stress limits, which corresponds corresponds toa to a circumferential circumferential crack extent of approximately 292 degrees or 81% (Reference apRroximately (Reference 18). The failure definition definition is coniservative since the threshold ligament coKServative ligament is based based on satisfying primary stress limits using a safety factor of 3 (and 1.5 for emergency emergency and faulted conditions).

The failure definition definition also does not take credit for the Technical Technical Specification Specification required 1 gpm leak detection capability, which as described in Appendix A may occur at a somewhat somewhat smaller crack extent depending depending upon the radial clearance clearance in the penetration penetration annulus. A conservative conservative failure definition Is appropriate for this is appropriate this risk assessment considering assessment conSidering the current weakness current weakness in industry understanding of OD 00 PWSCC and because because it may bound uncertainties uncertainties inherent in the probabilistic probabilistic fracture mechanics mechanics data.

of 56 Page 30 0156 NRC004-1201

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._. NON-PROPRIETARY 51-5012567-01 51-5012567-01 RV RV Head Nozzle and Head Nozzle and Weld Weld Safety Safety Assessment Assessment The The results of the Monte Carlo the.Monte Carlo simulation simulation are are illustrated illustrated by by the the histograms histograms shown shown in in Figures and 8. Based on the Monte Carlo simulatlon,an aD-initiated crack Figures 7 and 8. Based on the Monte Carlo simulation, an OD-initiated crack above above the J-grooveJ-groove weld weld wouldwould take take aa meanmean time time of of 8.9 years to grow 8.9 years grow toto aa through-wall through-wall state, state. andand aa mean mean time of of 28 28 years years to to result result in in nozzle nOZZle failure failure (or (or LOCA)

LOCA) due due to net-section net-section stress. stress. For For comparison, comparison, 97.5% 97.5% of the the time-to-failure time-to-failure distribution (non-parametric) distribution (non-parametric) is greater greater than than the the point point estimate estimate reported reported in Section Section 8.3.2 years for 3.3.2 (7.5 years for aa crack crack to reachreach 75% circumferential extent).

75% circumferential extent). ThisThis reflects the reflects the conservatism conservatism that that isis inherent inherent in in the the deterministic deterministic approach.

approach.

The The time-to-failure time-to-failure histogram histogram (Figure (Figure 8) has has been been partitioned partitioned into into two-year two-year probability Increments to probability increments to correspond correspond to the the worst-case visual Inspection intervals worst-case visual inspection intervals for for B&WOG B&WOG plants plants (i.e.,

(Le., plants plants withwith a two-year two-year fuel fuel cycle).

cycle). The The table table (see Figure Figure 88 inset) inset) shows shows the the probability probability that that thethe OD00 crack crack will grow grow to to failure failure within within thethe time time indicated indicated assuming assuming there there is no no detection detection by by visual visual inspection inspection (boron (boron crystals).

crystals).

The The opportunities opportunities for for detection detection will will be be added added at at two-year two-year intervals intervals in in the the event event tree quantification tree quantification (see Figure Figure 5).

9.3.5 9.3.5 Probability Probability of of Core Core Damage Damage The most The most likely likely consequence consequence of of CRDM CRDMnozzie nozzle failure failure (critical (critical size size crack) crack) isis leakage leakage that that is is within within the the capacity capacity of o! the the makeup makeup system. system. If a complete complete severance severance of of the the CRDM CROM nozzle nozzle occurs, occurs,the the break break size size will will be be within within the the range range ofof what what mostmost B&WOGB&WOG PRAs PRAs identify identify as as 'aa medium medium break break LOCA.LOCA. However,However, aa smaller break size smaller break size could could result result if there is if there isa a partial partial failure failure of of the the nozzle.

nozzle.

The The conditional conditional probability probability of of core core damage damage given given aa small-small- or or medium-sized medium-sized .LOCA LOCA can can be be readily readily determined determined from the the plant-specific plant-specific B&WOG B&WOG PRAs. PRAs. In Ina a B&WOG B&WOG PRA, PAA, the conditional core the conditional core damage probability (CCDP) damage probability (CCDP) for for aa medium medium break break LOCA LOCA is on average worse than is on average worse than for a small break LOCA. for a small break LOCA. Therefore, Therefore, as as' aa representative value, the risk assessment uses the representative value, the risk assessment uses the average CCOP for a medium- average CCDP for a medium-4x10-33 break LOCA break LOCA from survey of from aa survey of the B&WOG PRAs, theB&WOG PRAs, which which is is approximately approximately 4xl0-(Reference (Reference 35). 35). UseUse of of this this CCDP CCOP is is conservative conservative because because plant plant mitigation mitigation response response will be better for a break at the top of the vessel than for will be better for a break at the top of the vessel than for the the LOCAs LOCAs typically typically considered considered in in the the PRAs PRAs (see (see Section Section 8.0).

8.0).

9.3.6 9.3.6 RiskRisk Results Results forfor ODOOPWSCCPWSCC The The estimated estimated frequency frequency and and~r.ibabilities R6babilities in in the the preceding preceding sections sections are are used used to to quantify the event tree shown in Figure 5. The event quantity the event tree shown in Figure 5. The event tree shows the progression tree shows the progression of of sequences sequences startingstarting withwith thethe initiating initiating event event "CRDM "CROM leaks." leaks." Each Each sequence sequence can can result in success (e.g., no core damage) or failure/core damage. as noted result in success (e.g., no core damage) or failure/core damage, as noted by by the the "S" and the "CD" in the "Successor Core Damage" column. One sequence is "S" and the KCD" in the "Success or Core Damage" column. One sequence is Page 31 Page 31 of of 56 56 NRC004-1202 NRC004-1202

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• • • • NON-PROPRIETARY 51-5012567-01 RV Head Nozzle and Weld Safety Safety Assessment Assessment identified as "CO"CD Residual," recognizing that inspection and crack growth may continue beyond the eight years explicitly explicitly modeled modeled in the event*

event tree. The The contribution from these residual sequences sequences is not significant. In the event tree, at each decision point (success or failure), the conditional conditional failure probability (as estimated in In the previous sections) is shown. Multiplying Multiplying the appropriate appropriate branch branch failure probabilities results in the frequency frequency of core damage each sequence.

damage for each sequence.

Only sequences that result in core damage damage are quantified. When summed, the the sequence frequencies sequence frequencies provide an estimate estimate of the CDF due to ODPWSCC OD PWSCC of the the CRDM nozzles, which which has aa mean value of 3.4x1 0.7 per reactor-year.

3.4x1 07 Uncertainty Uncertainty in these results has has been addressed via the use of conservative conservative assumptions assumptions and the. the use of bounding bounding datadata inputs for the probabilistic fracture fracture mechanics. In particular, bounding assumptions were made mechanics. made for crack initiation initiation time, initial flaw distribution, distribution, and multiple crack initiation initiation sites. The crack crack growth rates used appear appear to produce produce results consistent consistent with the plant observations observations of TW cracks.

cracks. Other conservatisms conservatisms include the human error probability for visual include inspections, nozzle failure definition, and LOCA mitigation failure probability.

LOCAmitigation Therefore, it is concluded that the CDF results produced by this risk assessment assessment are reasonable in light of the limited industry knowledge knowledge base base for this failure failure mechanism.

The estimated core damage damage frequency (3.4x1 0"7 per reactor-year) compares frequency (3.4x10* compares acceptance guidelines favorably to the risk acceptance guidelines contained contained in Regulatory Regulatory Guide 1.174 (Reference 36) for core damage damage frequenCy.

frequency. Per these guidelines, the risk of operation with potentially undiscovered undiscovered CROMnozzie CRDM nozzle cracks Is is categorized categorized as as "very small." Regulatory Regulatory Guide 1.1741.174 also has accepta,,ce acceptal'ce guidelines for large early early release release frequency (LERF). The effect of the nozzle frequency (LERF). nozzle cracks cracks upon LERF is insignificant because the containment safeguards insignificant because safeguards systems not affected by systems are notaftected CRDM CRDM nozzle nozzle cracking.

cracking. The reactor vessel missile shields preclude preclude consequential consequential damage to the containment building in the unlikely event of CRDM nozzle nozzle detachment.

detachment. No other collateral collateral damage damage has has been been identified that may affect affect containment safeguards systems. Therefore containment safeguards Therefore it is concluded concluded that the risk associated associated with CRDM nozzle cracking at B&WOG plants is small and consistent cracking atB&WOG consistent Commission's Safety Goal Policy.

with the Commission's The public health risk associated.

assoCiated* with the CRDM nozzle nozzle cracking cracking is correspondingly very small.

correspondingly small. For example, the conditional conditional population population dose for aa medium medium break LOCA LOeA core damage accident at a typical B&WOGplant B&WOG plant (ONS) is 1.1 E4 person-rem

~erson-rem (Reference (Reference 37). For the estimated core damage frequency of 3.4xlC 3.4x10,7 per reactor-year, this corresponds corresponds to aa public health risk of only 3.7x10-4 3.7x1 04 person-rem/reactor-year, pers0r:-a-remlreactor-year, which is insignificant.

insignificant.

According Regulatory Guide 1.174, risk insights should be considered According to ReguJatory considered in an integrated fashion with traditional deterministic integrated evaluations (such as those deterministic evaluations those discussed in Sections 1 through 8). 8). The deterministic deterministic and risk evaluations taken Page 32 Page 32 of of 56 56 NRC004-1203 NRC004-1203

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• • • • 'NON-PROPRIETARY "-*- 51-5012567-01 RV Head Nozzle Nozzle and Weld Safety Assessment Assessment indicate that safety margins together indicate margins and defense-in-depth defense-in-depth are not significantly significantly affected affected by the CRDM nozzle nozzle cracking.

cracking. With effective visual Inspections;'

inspections, the the nozzle cra~king nozzle cracking does not significantly significantly increase increase the LOCA frequencyfrequency that is in B&WOG PRAs, nor does it increase assumed inB&WOG increase the frequency above the level that is assumed for design basis accidents. The consequences of a CRDM nozzle failure are less severe than the LOCAs assumed nozzle assumed in the FSAR analyses.

Also, the CRDM nozzle theCRDM nozzle cracking cracking has no effect on core damage damage mitigation, safeguards, or emergency containment safeguarcfs,or emergency planning effectiveness. Therefore, this effectiveness. Therefore, this risk analysis concludes concludes that the risk to the public due to CRDM nozzle cracking cracking Isis acceptable. The risk analysis analysis also supports the findings of the deterministicdeterministic analyses, which is that visual inspections of the RV head discover signs of head will discover CRDM nozzle CRDM nozzle leakage before there is a significant leakage before significant likelihood of total failure of a CRDM nozzle nozzle due to PWSCC.

SummarY and Conclusions 10.0 Summary Conclusions A safety assessment has beeh performed to address the potential been performed potential for PWSCC PWSCC

'cracking

'cracking of RV head penetration nozzles penetration nozZles and welds at the B&WOG plants. It addresses addresses both axially axially and circumferentially oriented circumferentlallyoriented flaws that have been observed in the Alloy 600 CRDM nozzles nozzles as well as axial/radial axial/radial flaws observed in the Alloy 182 182 J-groove partial penetration J*groove partial penetration welds used to attach Alloy 600 CRDM AJloy600 nozzles nozzles to low alloy steel RV heads.

heads. This safety safety assessment utilizes and builds builds upon the existing existing analyses performed performed for CRDMCRDM nozzle nozzle PWSCC (References (References 3, 7, and 8).

The results of detailed detailed stress analysis of the nozzle and weld regions of the RV head demonstrate that the circumferential" head demonstrate circumferential, or hoop, stress is generally higher generally higher than the axial stress at the same same location. On the downhill side of the noZZle, nozzle, the ratio of hoop stress to axial axial stress is about 211,2/1, and on the uphill side it is about 3/2. In the weld region, hoop hoop stresses are about about two times the axial stress at the same location.

location. It can therefore be concluded that if PWSCC cracking were to occur, flaws would predominantly predominantly be oriented oriented in aa longitudinal, longitudinal, or axial, plane, and as such would not promote catastrophiccatastrophic failure of the nozzle by ejection.

ejection.

Based Based on laboratory test data data for Alloy 182 weld metal metal .inin a PWR*

PWR environment, environment, crack growth growth through the J-groove J1700ve weld could occur rapidly (i.e., within one or or two years). Although continued crack growth into the low alloy steet steel head would expected due to the low not be expected tow susceptibility of this material to SCC. SCC, flaws in the the weld metal could continue to grow into the Alloy 600CRDM 600 CRDM nozzle, as seen at ONS-1 ONS-1 andand ONS-3.

ONS-3. It has been predicted that it would take almost four years for an axial OD 00 nozzle nozzle flaw to grow through-wall to the inside surface. At this point, a leak leak path into into .the the annular annular region between the nozzle and head head could be be present, depending depending on the location of the original flaw in the nozzle.

~.;,.

.~

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Nozzle and Weld Safety Assessment RV Head Nozzle Assessment 51-5012567-01 Any circumferential flaw above above the weld on the outside surface surface of the nozzle nozzle considered a safety concem.

should not be considered concern. A short, isolated flaw would take more than 10 years to grow through-wall, through-wall, while a long circumferential circumferential (where multiple flaws have joined) could grow from the outside outside surface surface to the Inside inside surface in about 3.5 years. In neither .case surface case would the structural integrity integrity of the the nozzle be compromised compromised to the point that the nozzle nozzle would fail by ejection.

Circumferential cracking Circumferential cracking has also been observed on the outside surface surface of CRDM nozzles at ONS-3, at the toe of the fillet weld that forms part of the structural attachment attachment to the reactor vessel head. Since these cracks are located at or below the weld, and not In the reactorreactor coolant pressure boundary, boundary, they are consi~ered to bea not considered be a safety concern concern from the standpoint of gross structural failure or release of radioactive water. Cue Due to the proximity of associated associated through-wall through-wall cracking below the weld, however, there thereisa is a concern that a through-wall circumferential circumferential crack could link up with two .ormoreor more through-wall axial cracks cracks and form a loose part.

Based on experience experience at ONS-3, circumferential circumferential and axial cracking cracking below the the weld is accompanied accompanied by through-wall axial cracking at and above the weld, weld, as as evidenced evidenced by by* deposits of boric acid crystals on the top of the head. It is concluded concluded from these results and observations that detectable detectable leakage would would precede precede the development development of ofaa loose part.

Concerns relating to aalack lack of fusion type tyle weld defect defect between the nozzle nozzle and weld have have been addressed addressed by considering the growth of a postulated 'Wrap- "wrap-arouncr circumferential around" circumferential flaw along'the cylindrical surface at the nozzle-to-weld along 'thecylindricaJ nozzle-ta-weld interface.

interface. Utilizing between the nozzle Utilizing radial stresses between nozzle and weld and PWSCC PWSCC crack growth rates for Alloy 182 weld metal, metal, it has been been calculated calculated that it would would take two years for a 17.5% wrap-around wrap-around flaw to grow grow to an allowable allowable 67% flaw size.

It has also been demonstrated demonstrated by a detailed stress analysis analysis that annular gaps gaps develop between the CRoM CRDM nozzle and the RV head head in the RV head penetration penetration of the B&WOG B&WOG plants. In the event of a through-wall through-wall crack in the J-groove weld weld or the portion of the CRDM nozzle in the annulus, these gaps form the natural leakage leakage path for the ReS RCS coolant to the 00 OD of the RV head. Assuming Assuming aa designed designed 0.0010 inch inch nominal nominal diametric interference, interference. the minimum minimum calculated calculated radial gap is 0.001 inch for both the center noZlie nozzle and the outermost nozzle outermost nozzle designs. The average average or representative representative radial gaps for the center nozzle nozzle and the the outermost nozzle are 0.0016 inch, and 0.002 inch. respectively.

inch, respectively.

Axial flaws are anticipated to be predominant predominant at both thelCand the ID and 00 OD of the CRDM nozzle based on tha the magnitude of the hoop stresses, although circumferential aftha circumferential envisioned on the 00 flaws can be envisioned OD and-have and* have been observed observed (both above and and below the weld).

weld). Axial flaws within the J-groove theJ-groove weld are also the most plausible plausible

~ .......

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. ~. Pg.. 34 o5 Page 34 of 56 Page 34 of 56 NRC004-1205 NRC004-1205

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.... NON-PROPRIETARY - " - 51-5012567-01 51-5012567-01 RV Head Head Nozzle Nozzle and Weld Safety Safety Assessment Assessment flaws due to high hoophpop stresses. These types of cracks are envisioned envisioned to break surface as pinhole type cracks or as tight PWSCC the surface PWSCC cracks. These tight fight cracks cracks would result in very low leakage rates as evidenced lOW leakage evidenced by the low volume of boric boric acid crystals cryStals found in the vicinity of CRDM nozzlesnozzles at ONS-1, ONS-1, ONS-2, ONS-3, and ANO-1.

ANO-1. It was estimated approximately 0.5 in3 was present around estimated that approximately around CRDM nozzle number 21 at ONS-1. ONS-1. However, However. observable leakage leakage is expected expected to occur well before crack propagation propagation would reach ASMECode ASME Code timits.

limits.

It has been shown that, assuming assuming a large portion of the nozzle nozzle cross-section cross-section contains a through-wall circumferential cack, through-wall circumferential crack, there is ample room for leakage leakage to occur occur before approaching the net section ligament. This will allow a section limit ligament leakage of steam through this large crack, thereby providing detectable leakage ample providing ample warning wamingto to prevent prevent the failure of the nozzle. In addition, evidence indicates indicates that nozzles are in an oval shape due to interaction the nozzles interaction with the closure head deformation. Therefore, there are gaps between deformation. between the nozzle nozzle and the headhead that will provide sufficient leak paths for foraa fairly large volume of steam to escape volume *of escape thereby providing thereby providing leak detection.

detection.

The allowable allowable lack of fusion size was previously affusion determined to be 67%,

previously determined 67%,or or 8.4 8.4 inches of circumferential circumferential extent. Also, the critical Jack lack of fusion size was was determined to be 85%

determined 85% or 10.6 Inches of the circumference. The leakage leakage rates rates were predicted using the same methodology as used for the evaluation of the predicted using axial crack. It was determined determined that a crack crask length length of 7.5 inches is required for the the center center nozzle design design to achieve achieve a 1 gpm leak rate. Similarly, it was established that a crack length of 5.0 inches is required for the outermost nozzle design design to achieve achieve a leak rate of 1 gpm. Since these cracks are less than the allowable allowable lack crack length of 8.4 inches, it ~s of fusion crack ;s concluded that these types of crackscracks will be detected detected by the plant's leak detection detection capability.

capability.

Boric acid corrosion corrosion concerns concems were addressed variety of conditions and addressed for a variety leakage rates potentially potentially assumed to occur. It was determined determined that corrosion of the RV RV head head penetration, penetration, at a maximum maximum volumetric metal loss rate of 1.07 In3/yr 1.07 in /yr would be possible. Various defect profiles profiles were postulated to model this level of corrosion corrosion for a-time a-time period of six years. It was concluded concluded that safe operation operation of the plant would not be affected as a result of this level of corrosion corrosion and that within this time, the leak will be detected detected during during a walk-down inspection the RV inspection of theRV head area.

All of the observed through-wall through-wall CRDM cracks in the B&WOG plants have been traced to origination in the vicinity of the weld and not at the end of CRDM nozzle. Failures in the end of the nozzle have have the potential to generate generate loose parts that could relocate within the RCS and compromise equipment operation or fuel-clad fuel-clad barrier integrity. Given Given the current knowledge knowledge of the residual stresses in the CRDM nozzles, FRA-ANPFRA-ANP has concluded that the through-wall axial cracks has concluded cracks present below below the weld initiate at the toe of the weld. These cracks cracks are not not Page Page 35 ofof 56 56 NRC004-1206 NRC004-1206

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- 51-5012567-01 RV Head Head Nozzle Nozzle and Weld Safety Assessment Assessment expected to propagate propagate to the pofnt point that a loose part will be generated before generated before some leakage is visible. Therefore, Therefore, all aspects of the CRDM CRCM cracks have been considered from aa safety analysis considered analysis perspective. This review has concluded concluded that simultaneous CRDM nozzles will not fail and that the failure of a single simultaneous multiple CRCMnozzles single CROM nozzle CRDM bounded by both the LOCA and non-LOCA plant analyses nozzle is bounded analyses completed to support current already completed already aJrrent plant operation.

evaluation was performed A loose part evaluation performed to evaluate the potential potential for loose parts from a failed CRDM nozzle to potentially potentially enter a control rod guide tube and and prevent the control control rod assembly assembly from being being fully inserted. It was concluded concluded that there was at least least a 25 percent chance of a loose part entering the guide tube pereent chance tube successful operation of that assembly. The LOCA and and potentially impairing successful and non-LOCA analyses assume that the control rod of highest non-lOCA highest worth is stuck out of the core. In addition, only a fraction of the remaining remaining worth is used in demonstrating that at least demonstrating least a 11 percent shutdown percent shutdown margin exists at hot zero power conditions. .

demonstrated through risk analysis that the risk from potentially It has been demonstrated potentially undetected CRDM nozzle undetected nozzle cracks is "very small" per the guidelines guidelines of Regulatory Regulatory Guide 1.174. The estimated core damage Guide damage frequency due due to 00OD PWSCC PWSCCofthe of the CRDM nozzles is 3.4x10*

CADM 3.4xl 0'7 per reactor-year. Conservative Conservative assumptions assumptions are made made in the risk assessment assessment to address uncertainty in the estimates of human address uncertainty reliability, probabilistic fracture reliability. probabilistic mitigation response. Taken mechanicS, and plant mitigation fracture mechanic$.

together with the results of the deterministic deterministic analyses, the risk analysis analysis demonstrates that visual inspections of the reactor vessel head will demonstrates will be sufficient sufficient minimize public to minimize Inspections will discover public risk. The visual Inspections discover signs of CRDM CADM nozzle or penetration nozzle penetration weld leakage before before there is a Significant significant likelihood that thethe leakage leakage will causecause CRDM CADM nozzle nozzle structural failure or detachment outside detachment due to outside diameter PWSCC.

diameter Finally, evidence to date suggests that it will require several years for the Finally. all evidence the material to degrade to the point that total failure of the component materIal component could occur.

During that time, if if a crack should form,fonn, leakage primary coolant on to the RV leakage of primary identified through routine visual inspections.

head can be identified inspections. The component component can repaired and retumed to service without jeopardizing then be repaired jeopardizing the health and safety of the p public.

u b l i c . ...

As Asa a result of the previously described evaluations performed described activities and evaluations performed by the B&WOG,B&WOG, the following conclusions conclusions have been reached regarding reached regarding degradation degradation of CRDM nozzles, nozzles, and RV head thermocouple nozzles.

nozzles. thermocouple attachment head attachment B&WOG plants:

welds at B&WOGplants:

1) B&WOG plant safety evaluation The B&WOGplant (Reference 3) remains valid.

evaluation (Aeference Page 36 of 56 NRC004-1207 NRC004-1207

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2) The B&WOG utilities comply with 10CFR50.55a 10CFR50.55a and continue to meet the intent of General Design Criteria Criteria 14, 30, 31, and 32 of 30,.31, Appendix A of 10CFR50.

10CFR50.

3) The potential potential for the B&WOG B&WOG plants sulfur-induced IGA or plants to have sulfur-induced sec SCC of CROM and thermocouple of CRDM thermocouple nozzles (Reference nozzles is very low (Reference 5).

4) The risk to the public due to CRDM CROM nozzlenozzle cracking is "very small" acceptable per the guidelines and acceptable Regulatory Guide guidelines of Regulatory Guide 1.174.

5) Visual inspections of the reactor vessel head will discover discover signs of CRDM nozzle leakage CROM leakage before significant likelihood of total before there is a significant total failure faHure of aa CROM CRDM nozzle due to PWSCC.

6) Inspections, other than visual examinations Inspections. examinations In accordance with GL in accordance 88-05, are not necessary necessary from a safety perspective.perspective.

7)

7) One of the most susceptible B&WOG plants. plants, ONS-2, has Inspected Inspected all 69CROM 69 CRDM nozzles nozzles Inin 1994 and two follow-up inspections inspections on the the nozzles nozzles identified with flaw-like indications.

indications. Recent Observations at Recent observations ONS-1, ONS-2, ONS-3, and ANO-1 have ONS-1, have also added credence credence to the safety assessments assessments that have been performed. performed.

8) 8} All A" B&WOG utilities continue to perform perform visual inspections of the the head in accordance RV head accordance with their respective respective Generic Letter 88-05 Gene/'icLetter and Bulletin 2001-01 responses.

9) The B&WOG will continue continue to share B&WOG plant inspectiOn shareB&WOG inspection data data participate in agreed and participate agreed upon joint Owners Owners Group (e.g., MRP)

(e.g.* MAP) activities with the U.S. nuclear nuclear industry on this issue. issue..

10) The B&WOG B&WOG will continue to monitor this issue.

11.0 References References

1) Evaluation of the Potential "Safety Evaluation Potential for and Consequences of Reactor Reactor Head Penetration Vessel Head Penetration Alloy 600 10 ID Initiated Initiated Penetration Penetration Cracking,"

CEN-607, May 1993.

2) "Alloy 600 600* Reactor Vessel Head Head Adapter Adapter Tube Cracking Safety WCAP-13565*. Rev, Evaluation," WCAP-13565. Rev. 1f F~bruary February 1993 1993..

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3) "Safety Evaluation "Safety Evaluation for for B&W-Designed 8&W-Deslgned Reactor Vessel Head Reactor Vessel Head Control Control Rod Rod Drive Mechanism Nozzle Cracking;"

Mechanism Nozzle Cracking," BAW-10190P BAW-10190P (B&W Owners Owners Group Group Proprietary),

Proprietary), May 1993. .

4) Generic Letter 97-01: "Degradation Genedc "Degradation of Control Control Rod Drive Mechanism Mechanism Nozzle And Other Vessel Closure Head Penetrations:

Nozzle Penetrations,' U.S. Nuclear Nuclear Regulatory Regulatory Commission, Office Offlceof of Nuclear Nuclear Reactor Regulation, Washington, DC, Apdl1, April 1, 1997. .

5)

5) "B&WOG Integrated Response to Generic "B&WOGlntegratedResponse Generic Letter 97-01: 'Degradation

'Degradation of Control Rod Drive Mechanism Nozzle And Other Vessel Mechanism Nozzle Vessel Closure Head Penetrations'," BAW-2301, BAW-2301, July 1997.

6) Title 10 of the Code of Federal Regulations.

Regulations, Part 50, U.S. Nuclear Nuclear RegulatorY Regulatory Commission, Washington, DC. .

7) "External Circumferential Circumferential Crack Crack Growth Analysis for B&W-Design B&W-Design Reactor Vessel Vessel Head Control Rod Drive MechanismMechanism Nozzles," BAW-10190P, BAW-10190P.

Addendum Addendum 1 (B&W Owners Group Proprietary), :December December 1993.

8) "Safety Evaluation for Control Rod Drive Mechanism Mechanism Nozzle J-Groove J-Groove Weld," BAW-10190P, Addendum Weld,* Addendum 2 (B&W Owners Owners Group Proprietary),

Group Proprietary),

December December 1 1997.9 9 7 " .

9) Generic Letter No. 88-05, Generic 88-QS, "Boric Acid Corrosion Corrosion of Carbon Steel Reactor Reactor Pressure Boundary C"rnponents Pressure Boundary C,'-mponents in PWR PWR Plants," U,S,U.S. Nuclear Regulatory Regulatory Commission, March March 17, 1988.

17,1988.

10) Bulletin Bulletin 2001-01: "Circumferential "Circumferential Cracking Cracking of Reactor Reactor Pressure Vessel Head Head Penetration Penetration Nozzles," US.

U.S. Nuclear Regulatory Commission, August Regulatory CommiSsion, August 3,2001.

3,2001.

11)

11) "Oconee Nuclear Station Units Nuclear Station Units 1, 2, &

& 3, Docket Nos, 50-269, 270, and Nos. 50-269, and 287, Response Response to NRC Bulletin 2001-01: Circumferential Circumferential Cracking Cracking of Reactor Pressure Pressure Vessel Head Penetration Nozzles,"

Head Penetration Nozzles,' Duke Energy letter letter from W.R W.R. McCollum McCollum to U.S. Nuclear Nuclear Regulatory Regulatory Commission, August 28, 2001.

2001.

. *~ . . **i.~".*

12) "Arkansas Nuclear Nuclear One One - Unit 1.,1, Docket No. 50-313, Ucense No. DPR-51, 50-313, U~ense DPR-51, 30 Day Response to NRC Bulletin 2001-01 for ANO-1; AN0-1; Circumferential Cracking of VHP Nozzles," Entergy letter from C. Anderson to U.S.

Entergy letter Nuclear Regulatory Regulatory Commission, September September 4, 2001 (1CAN090102).

(1CAN090102).

13)

13) "Crystal River UnitUnit 3 - Response to NRC Bulletin Bulletin 2001-01, 2001-01, Cracking of Reactor "Circumferential Cracking Pressure Vessel Head Reactor Pressure Head Penetration Page 38 of 56 NRC004-1209 NRC004-1209

• • • • NON-PROPRIETARY NON-PROPRIETARY **** 51-5012567-01 RV Head Head Nozzle Nozzle and Weld Safety Safety Assessment Assessment Nozzles," Progress Energy Nozzles)" Energy letter from D.E. Young to U.S. Nuclear Nuclear Regulatory Commission, August 30, 2001 (3F0801-06).

Regulatory

14) Exelon/AmerGen Response to NRC Bulletin 2001-01, ExeloniAmerGen 2001-01, "Circumferential Cracking of Reactor Pressure Pressure Vessel Head Penetration Nozzles," letter Head Penetration from J.A. Benjamin Nuclear Regulatory Commission, Benjamin to U. S. Nuclear Commission, August 31.31, 2001 (R8-01-182).

(RS-01-182).

15) "Response to NRC Bulletin 2001-01, 2001-01, "Circumferential Cracking of Reactor Reactor Penetration Nozzles," letter fromG.G.

Pressure Vessel Head Penetration from G.G. Campbell to' to Regulatory Commission, U.S. Regulatory Commission, September 4, 2001.

4,2001.

16) "Crack Growth of Alloy 182 Weld Metal Metal in PWR Environments Environments (PWRMRP-(PWAMRP-

. 21)," TR-1000037, Bectric 21 )," 18-1000037, Electric Power Research Institute.

Institute, June 2000.

17) "Reactor "Aeactor Coolant System Pressure Boundary Boundary Leakage Leakage Due to to Cracks Cracks Found in Several Several Small Bore Reactor Vessel Vessel HeadHead Penetrations,"

Licensee Event EVent Report 2000-006-00, 1, Docket 2000-006-00, Oconee Nuclear Station 1,Docket Number 05000-269, 0500D-269, JanuaryJanuary 2, 2001.

2001.

18) "O-3 CAOM Killian, D.E., "OC-3 CRDM Nozzle Circumferential Circumferential Flaw Evaluations," 32-5012403-00, 5012403-00, April 2001 (FRA-ANP (FAA-ANP Proprietary).

19) Vreeland, D.C., "Corrosion of Garbon Vreeland, Carbon Steel and Low Alloy Steels in In Primary Systems Systems of Water-Cooled Water-Cooled Nuclear Reactors," Presented at Presented Netherlands-Norwegian Netherlands-Norwegian Reactor School, Kjeller, Norway, August 1963.

20) Uhlig, H.H., ed., Corrosion Handbook, Wiley, New York, 1948, 1948, p 125.

21)

21) Shvartz, G.C.,G.C., and Kristal, H.M.,H.M., Corrosion of Chemical Apparatus, Apparatus, Chapman Chapman and Hall, London. London, 1959, pp 53-70.

22) Moore, K.E., "Stress Corrosion Cracking Moore. K.E., Cracking of Low Alloy Steel," 51-5012047-

00. March 2001 (FRA-ANP Proprietary).

rub March 2001 (FRA-ANP Proprietary).

23) "Cracking in Charging Pump Casing Cladding," IE Information Information Notice Notice No.

8aD-38, Nuclear Nuclear Regulatory October 31, Aegulatory Commission, October 31, 1980.

24) "Boric Acid Corrosion Corrosion of Charging Charging Pump Casing Caused by Cladding Cladding Cracks,'" Ig Cracks," Information Notice 94-63, Nuclear IE Information Nuclear Regulatory Regulatory Commission, Commission, August 30, 1994. 1994.

25) Snow, F F.,.* "PT aPT Inspection Inspection Report Resolution,' 51-5011639-00, Report Resolution," 51-5011639-00. February 2001.

2001.

Page 39 of 56 NRC004.-1210 NRCOO4.-1210

• lk** NON-PROPRIETARY

• • • NON-PROPRIETARY lk*_

• 51*5012567"()1 51-5012567-01 RV Head Head Nozzle and Weld Safety Assessment Assessment

26) v .C. Summer Nuclear V.C. Nuc/ear Station Alpha Hot leg Leg Evaluation Evaluation and Repair, Presentation to the Nuclear Regulatory Commission Region II.

Presentation II, December December 20,2000.

27) "Boric Acid Corrosion Guidebook "Boric. Guidebook - Recommended Recommended Guidance Guidance for Addressing Addressing Boric Acid Corrosion Corrosion and leakage Leakage Reductionlssues,"TR-Reduction Issues," TR-Electric Power Research Institute, April 1995.

102748, Electric

28) Wiemer, J.A.,

Wierner, J.A., "Loose "LooseCRDM CRDM Nozzle Components 51-5012057,-

Components in RCS," 51-5012057-00, April QQ, April 2001.

2001. ...

29) Xu, Hongqing, Xu, Hongqing, *CHECWORKS "CHECWORKS RHNM RHNM PWSCC Risk Assessment," §1: 51-5013250-00, 5013250-00, June June 2001 (FRA-ANP Proprietary).

30) 30) CHECWORKS PWR Vessel and Intemals CHECWORKS Internals Application:

Application: RPV Head Nozzle Module, Version 1.0, User Guide," TR-103198-P8. TR-1 03198-P8 Final Report, December 1998, Electric Electric Power Power Research Institute, Palo Alto.

Alto, CA.

31)

31) Levinson, Levinson, S.H., S.H., "Probability that CRDM Leakage is Detected,"

Detected," 32-5013324-00 5013324-00. June 2001. 2001.

32) W. Hannaman, "Human Cognitive Reliability Reliability Model for PRA Analysis," RP-2847-1, Electric Power 2847-1, Electric Power Research Institute, December December 1984.

33) D. Swain and H. E. Guttman, "Handbook of Human-Reliability Analysis Human-Reliability Analysis with Emphasis on Nuclear Nuclear Power Plant Applications/Final ApplicationslFinaJ Report," Sandia Sandia National National Laboratories, prepared for the U. S. Nuclear Nuclear Regulatory Commission, SAND80-0200.

SAND80-0200, NUREGlCR-1278, NUREG/CR-1278, August 1983.

34) Mazurkiewicz, Mazurkiewicz, S.M., Carlo Evaluation S.M., "Monte Car/o Evaluation of Circumferential Circumferential Flaws Raws in B&WOG CRDM Nozzles," 32-5013346-01, S&WOG 32-5013346-01, August 2001 (FRA~ANP (FRA-ANP Proprietary). '

35) Enzinna, R.S. and Levinson, levinson, S.H.,

S.H., "S&WOO PSA Comparison, 47-UB&WOG PSA,Comparison,"!Z:

5006733-00, January-2000.

5006733-00, January.2000.

36) Regulatory Guide 1.174.

Regulatory 1.174 "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Plant-Specific Changes to the Risk-Informed Decisions on Plant-Specific the Licensing Basis," U.S. Nuclear licensing Regulatory Commission, Nuclear Regulatory Commission, July 1998.

37) Oconee Nuclear Station PRA, Revision Oconee Revision 2, Duke Energy Corporation, Energy Corporation.

December 1996.

December 1996.

38) "Leakage Assessment "leakage Through CRDM Nozzle and Closure Head," SAW-Assessment ThroughCRDM BAW-2213 (B&W Owners 21_3 Owners Group Group Proprietary),

Proprietary), June 1994.

1994*

., .iIhI!l.~ ". * ...*.*~.'1~.

Page 40 of 56 NRC004-1211 NRC004-1211

_. NON-PROPRIETARY

• • • • NON-PROPRIETARy .... " 51-5012567-01 51-5012567-01 RV Head Nozzle and Head Nozzle and WeldW.ld Safety Safety Assessment Assessment Figure 1. Side View Schematic Schematic of of B&W-Design S&W-Desrgn Reactor Vessel Head, CRDM Reactor Vessel Nozzles, Thermocouple Nozzles, Thermocouple Nozzles, Nozzles, and Insulation. Insulation.

RV HEAD INSULATION INSULATION SERVICE STRUCTURE SERVICE STRUCTURE cpnu r;

• ..tI II./111'1 Mnr771FV'F----

I II 1,.!/..,.I1... I..1,...*../

THERMOCOUPLE THERMOCOUPLE NOZZLE (ONS-1 AND TMI-1 (ONS-1 TMI-1 ONLY)

ONLY)

SUPPORT STEEL SUPPORT 8 OR 9 ACCESS HOLES HOLES II IN SERVICE STRUCTURE IN SERVICE STRUCTURE SUPPORT SUPPORT (ONS-1, ONS-2, (ONS-1, ONS-2.

ONS-3, CR-3, AND TMI-1 ONS-3, TMI-1 ONLY)

ONLY) 18 ACCESS OPENINGS 18 ACCESS OPENINGS 2" MIN GAP BETWEEN 2"

"MOUSE-HOLES" ALL INSULATION AND TOP INSULATION TOP "MOUSE-HOLES" ALL 8&WOG PLANTS B&WOG PLANTS OF RV HEAD OF RV HEAD Ii I

Note: The thennocouple thermocouple nozzles were removed from ON8-1 ONS-1 at EOC-19.

..t;~:5  :

Page 41 of 56 NRC004-1212 NRC004-1212

...... NON-PROPRIETARY NON*PROPRIETARY **** - 51-5012567-01 51-5012567-01 RV RV Head Head Nozzle Nozzleand and Weld Weld Safety SafetyAssessment Assessment Figure Figure2. 2. Plenum Plenum Cover CoverAssembly.

Assembly.

Top TopView View

~~~-- Cover CoverPlate Plate Keyway Keyway-~

SupportPad Pad BottomFlange Bottom Flange Support SupportRing Ring Segment Segment SupPOrtFlange Support Flange M fing Lug Ufting Lug Side SideView View

- Support Ring Segment Support Flange Weldment WeldmentRibs Ribs Page Page42 42of0156 56 NRC004-1213 NRC004-1213

~..*NON-PROPRIETARY NON-:PROPRIETARY **** 51-5012567-01 51-5012567-01 RV Head Nozzle RV Head Nozzleand andWeld WeldSafety SafetyAssessment Assessment Figure Figure3.3. Control ControlRod RodSpider SpiderAssembly.

Assembly.

Coupling Spider Spider ii Top TopView View F] L irn Neutron Absorbing Material Neutron Absolblng Material Control Rod control Rod l6

• to ~.

1,- ' *. 11 Page Page43 43ofof56 56 NRC004-1214 NRC004-1214

-* NON-PROPRIETARY **-

• • • • NON-PROPRIETARY **

RV Head Nozzle and Weld Safety Assessment Assessment 51-5012567-01 Figure 4. Control Rod Guide Brazement Assembly Guide Brazement Rod Guide Brazoment Guide Brazement Sectors Rod Guide Sectors Rod Guide Guide Tubes Tubes ROd Rod Guide Spacer Castings GuideSpacer Castings Rod Guide Sectors Sectors Rod Guide Tubes JIl]lr::~. ~4---""-:\ Spacer Casting Spacer Casting Page 44 of 56 NRC004-1215 NRC004-1215

0

• • • • NON-PROPRIETARY****

NON-PROPRIETARY **** 51-5012567-01 RVHead RV Head NozzleNozzle and and Weld Safety Safety Assessment Assessment Figure 5.

FipIure s. Event Eyent Tree Frequency of Core Tree for Freauency Damage from Core Damage from Outside PWSCC in Diameter PWSCC Outside Diameter B&WOG CRDM in B&WOG CRDM Nozzle Nozzle CROM LEAKS CRDM CRACK LEAKS CRACK, FIRST VISUAL CRACK FIRST VISUAL CRACK SECOND SECOND CRACK CRACK VISUAL CRACK THIRD VISUAL THIRD CRACK CORE CORE SUCCESS SUCCESS FREQUENCY FREQUENCV INITIATES INITIATES INSPECTION INSPECTION GROWS GROWS TO TO VISUAL VISUAL GROWS TO GROWS TO INSPECTION INSPECTION GROWS TO GROWS TO DAMAGE DAMAGE OR OR CORE CORE (per REACTOR-YR)

(per REACTOR.YR)

AND GROWS AND GROWS FAILS (2 FAILS (2 YRS .FAILURE ININ 2. INSPECTION 2 INSPECTION FAILURE IN FAILURE IN44 FAILS (6 YRS FAILS YRS FAILURE IN 6G OCCURS DUE FAILURE DUE DAMAGE DAMAGE FAILURE TO FAILURE OF BORON OP BORON TO 44 YEARS TO YEARS FAILSFAILS (4 (4 'fRS YRS TO. YEARS TO$ 'fEARS OF I.ORQN or aOROH rO.rrARS TO 4 YEARS TaLOCA TOLOCA IN cc2I VEARS YEARS CRYSTALS)

CRYSTALS) OF BORON OF BORON CRYSTALS)

CRYSTALS)

IN CRYSTALS)

CRYSTALS)

~

"l'ot S

1' 8

S 1.1.*'

ConUnu.

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6.59-11 Total 3.40-7 3.4.*7 CDF CDF z

a Page 45 of 56 oo o

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• • • • NON-PROPRIETARYARY ****

• • • • 51-5012567-01 RV Head Safety Assessment Head Nozzle and Weld Safety Assessment Jtlgure 6. Crack Growth Rate Assumed in Monte Carlo Simulation Figure 1.4 . . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . .

1.2+-----------~-------------------------------------------r--~~30 1.2 30

~ 1.0 25 as CD -.:-

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00 t.. ..

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0% 10%

10% 20% 300%

30% 40% 50% 60% 70% 80% 90%

90%'

Through-Wall Circumferential Through-Wall Extent Circumferential Crack Extent Z

0 o

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.......... NON-PROPRIETARY NON-PROPRIETARY **** .*.* 51-5012567-01 51-5012567-01 RV Head RV Head Nozzle Nozzleand and Weld Safety Assessment Weld Safety Assessment Ffeure Figure7. Probability of

1. Probability Through. Wall Crack ofThrough-Wall versus Time Crack versus after Initiation Time after Initiation of Outside Diameter ofOutside Diameter PWSCC PWSCC 16000 16000 Sampics Samples = = 80,000 80,000 First First TW TW '" 1.56 1.56 years yoars I..

Mena Mean = = 8.92 years 8.92 years 14000 14000 Median Median == 7.22 years 7.22 ycars prghnhjlUy of Through-Wail 12000 12000 (It tl <<22 years 3.39x10"3 3.39x I0.3

.~

~ 10000 10000 *2years years

• 2ycars << tl <4

< 4 years years 1.18x10' 1.18xI0*'

1.7 !x 10"

-~

0(Io 2.

8000 8000 ~'.::: ~f:.\ :(::' ',' .

44 years < I < 6 years ycars < l < 6 years 6 years < t < 8 years 6 years <t < 8 years 1.71,,10*'

1.46x 10" 1.46xlO"

.c * " ':.'

9en .~:".:':~.

I e

.c 6000 6000 4000 4000 2000 2000

1

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,C;,Tim (years)ý '

Tim e (years) z

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• • • • NON-PROPRIETARY****

NON-PROPRIETARY **** 51-5012567-01 51';5012567-01 RV Nozzleand HeadNozzle FlVHead and Weld SafetyAssessment WeldSafety Assessment Probabilityof Figure8.8. Probability Figure Net-SectionFailure ofNet-Section Timeafter versusTime Failureversus Initiation of afterlnitiation DiameterPWS!CC OutsideDiameter ofOutside pwsCC 4500 4500 Samples = 80,000 FirJit Failure :: 3.56 years Mean= 27.95 yean 4000 4000 Median = 23.22 years 3500 3500 Pmhqbiljtx pf fRj"TS 3000 3000 " t < 2 years <1.25xIO*$

2 years<l < 4 ycars 1.50x10""'

92500 2500

!:;, 4 years < l < 6 years 6 years < t < 8 ycars 8; JhlO*l

~

,, 2000 2000 1500 1500 1000 1000 I1I* I 500 500 o0I Q N

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Z Al o

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• • • • NON-PROPRIETARy*

NON-PROPRIETARY ...

• 51-5012567-01 RV. Nozzle and Weld Safety Assessment RV Head Nozzle.and ....

Appendix Appendix A

{.

Leakage Assessments Assessments r-. . .. ' .

Page 49 of 56 NRC004-1220 NRC004-1220

• • • NON-PROPRIETARY NON-PROPRIETARY **** **** 51-5012567-01 51-5012567-01 RV RV Head Head Nozzle Nozzle and W.,ldWeld Safety Safety Assessment Assessment The leakage assessments due leakage assessments due to various various postulated postulated flaws flaws andand due due to lack of fusion fusion in in the CRDM nozzle/J-groove weld region CRDM nozzle/J-groove addressed in this region are addressed this section.

section.

As As aa result result of stress stress analyses analyses of of the

the B&W-design B&W-design CRDM nozzles. itit has CRDM nozzles, has been demonstrated that during normal previously demonstrated previously normal operation steady state conditions operation steady conditions an annular annular gapgap develops develops (above the CRDM CROM weld weld to the RV head) between between the the CRDM CRDM nozzle nozzle and and the RV head penetration (Reference head penetration (Reference 3). 3). Of particular interest particular interest is the prediction prediction of a radial gap gap in a previously previously interference-fit interference-fit region. The The prediction prediction of this radial gapgap during steady steady state state operating operating conditions utilized in conditions is utilized the assessment leakage rates through assessment of leakage through the CRDM nozzle/head nozzle/head annulus.

The The radial gaps are different for the two types of CRDM are different CRDM nozzlesnozzles that were were evaluated by aa detailed stress analysis, evaluated analysis. the center center nozzle nozzle design and the th$

outermost nozzle design. For the center nozzle design. center nozzle, nozzle, thethe Initial Initial interference-fit interference-fit between the nozzle and the head betw$enthe separates to form a 0.003 inch head separates inch maximum radial gap above the weld during steady steady state conditions.

conditions. The averageaverage radial gap gap is 0.0016 inch and the minimum radial gap is 0.001 inch minimum radial illustrated in Figure inch as illustrated Figure A-1.

A-1.

For the outermost nozzle, the radial clearanceclearance in the initial interference fit region initial interference approximately 0.001-inch is approximately Is minimum during steady state conditions 0.001-lnch minimum conditions as depicted depicted in Figure Figure A-2. However, a major major portion of the periphery periphery of the CADM CRDM nozzle/RV nozzletRV head penetration penetration shows clearance of at least 0.002 inch and the shows a radial clearance the maximum radial gap is about 0.003 inch. inch.

A.1 A.l Flaws in CRDM.NozzJe Axial Raws CRDM Nozzle Aboye Abo,ve the J-Groove Weld Weld Leakage assessments Leakage assesSments for postulated through-wall axial flaws in the CRDM postulated through-wall nozzle above the J-groove weld were previously nozzle previously addressed Reference 3 and addressed In Reference summarized below.

summarized Reactor coolant system (ROS)

Reactor (RGS) leakage rates through postulatedpostulated CRDM nozzle nozzle cle~lrances between cracks and the annulus clearances between the nozzle nozzle and reactor reactor vessel head were predicted by by a parametric analysis. Both the crack lengths and annulus annulus clearances were varied. Because clearances pressure-high energy Because of the high pressure-high energy conditions conditions in RCS, the sub-cooled coolant saturates, flashes, the ReS. flashes. and then chokes at the exit of either the crack or annulus.

Leakage rates were obtained through an iterative process process using the the Equilibrium Model (HEM) critical flow tables and by solving single Homogeneous Equilibrium Homogeneous single loss correlations. Since the flow chokes at either the and two-phase pressure *Josscorrelations. the crack/annulus interface) or at the exit of the annulus (i.e.,

exit of the crack (i.e., crack/annulus top of the penetration) for any given crack length and annulus clearance, both annulus clearance.

possibilities were considered considered in the analysis.

analysiS.

Therefore, the flow through the crack and annulus clearance was broken into into two separate leakage flow paths to account for the the two possibilities: (1) single and two-phase flow through the crack with choking choking at the exit of the crack, crack, and (2) (2)

Page 50 50 of 56 NRC004-1221

• • • • .NON-PROPRIETARY ****

-NON-PROPRIETARY **** 51-5012567-01 RV Head Head N9zzleand Nozzle and WeldWeld Safety Assessment Assessment single single phase flow through the crack and single and two-phase flow through through thethe clearance dearance annulus, with choking choking at the exit of the annulus.

For the first path with flow choking at the exit of the crack, the downstream leakage leakage paths were calculated. The path with the lesser flow rate was considered to have considered have the actu~

actual flow rate. Because Because of the choking properties of the flow, the greater greater flow rate was not possible. Thus, ifif the flow rate through the path with choking at the exit of the crack is Is less than that through the crack and the annulus, then the flow rate through the C(ack crack and the annulus is limited by choking at the exit of the crack.

In crack limited problems.

problems, the flow chokes at the crack exit. The pressure just upstream upstrearilofof the exit is assumed to be the exit pressure. Using Using this pressure, the the RCS enthalpy, and the HEM tables, a trial critical Res critical mass flux is established. This This flow rate is used in crack calculations to determine a new value for crack pressure loss calculations the exit pressure. When the assumed and calculated calculated values of the exit pressure the solution has converged agree, thesoJution converged and the crack limited flow rate is established.

The crack crack pressure pressure loss calculations are divided into two calculations: calculations: sub-cooled sub-cooled flow and two-phase flow.

The results of the analysis show that for annulus annulus clearances greater than 0.0001 dearances greater inch and crack crack lengths less than 3 inches, the limiting limiting factor is the size of the the crack, while in cracks cracks longer than 3 Inches, the flow does not reach saturation conditions in the crack and therefore citokes at the exit of the annulus.

therefore chokes annulus. For an an clearance less than 0.0008 inch.

annulus clearance inch, the flow rate will not exceed exceed 1 gpm Likewise, for a crack length regardless of crack size. Ukewis9,fora length of 2 inches inches and shorter, the "ow rate will not exceed leakage IIOW exceed 1 I gpm regardless of annulus clearance. clearance.

For a crack length of 2 inches and a maximum annulus .annulus clearance clearance of 0.003 inch, the leakage flow rate was determined determined to be beO.SS90.559 gpm. However, it was was demonstrated demonstrated that as the crack extends from 2 to 3 inches in length, the flow rate rate would approach approach and exceed exceed the leak detection capability detection capability rate of 1 gpm for annulus clearances clearances of 0.001 inch and greater.

In addition, an Independent Independent leakage leakage assessment was also performed performed as documented documented in Reference Reference 38 and summarized summarized below.

The objective of the report report was to demonstrate that sufficient sufficient leakage of primary primary coolant, beyond the 1 gpm leak detection detection capability per RegulatoryRegulatory Guide 1.45 1.45 requirements, is feasible if a PWSCC feasible if indication of sufficient PWSCC indication sufficient size occurs occurs in thethe CRDM nozzle. The evaluation evaluation was based on applicable industry leak test data data to the CRDM nozzle/head annulus CADM nozzJe/head annulus (subsequently written as "CRDM annulus") annulusj gap.

experimental data on two-phase An inventory of experimental experiments were two-phase critical flow experiments were reviewed to help identify those that are applicable to the problem of predicting reviewed predicting leakage rates through the CRDM CRDM nozzle and the annulus between between the nozzJe nozzle and the RV penetration.

Page 51 of 56 56 NRC004-1222 NRC004-1222

I

_. NON-PROPRIETARY

• • • • NON-PROPRIETARY _**** ... 51-5012567-01 51-5012567-01 RV Head Nozzle Nozzle and Weld ,Safety Safety Assessment Assessment Only the most pertinent data data from the literature of experimental experimental investigations investigations were considered considered in the assessment assessment of leakage rate through the CRDM annulus.

The experiments experiments were determined to be pertinent were detel'J1llned pertinent based on review against key thermal-hydraulic parameters thermal-hydraulic parameters for the evaluation evaJuationof of leakage through the CRDM nozzle/closure head annulus.

nozzle/closure head annulus.

pertinent data were identified in the work of Agostinelli.

The pertinent Agostinelli, etal.,

et al., Amos and Schrock, and Matsushima,et Matsushima, et al. (see Reference Reference 38 for these citings).

cI1lngs). The data data from the first two references, when related to the CROM trom CRDM problem predicted leakage rates greater leakage greater than 1 gpm. The data from the third reference, when when related to the CROMCRDM problem, corresponded corresponded to a leakage rate of 0.6 gpm.

0.6gpm.

However,

.However, the experiment was based on a stagnation stagnation pressure of only 975 psi pressure ofonty psi stagnation pressure associated with the CROM and the stagnation CRDM nozzle is 2250 psi.

Accounting Accounting for the higherhigher stagnation pressure should result in a predicted predicted leakage rate gre&ter than 1 leakager&.tegreE.terthan I gpm. Therefore, it is concluded In in the report that, based on the planfs plant's leak detection detection capability of 1 gpm within an hour per Regulatory Guide Regulatory Guide 1.45. 1.45, the leakage leakage* through the CRDM annulus (under (under thethe conditions discussed conditions discussed in the report) will be detectable. Furthermore, it should bebe noted noted that the prediction prediction of the leak rates rates given above above were conservatively conservatively determined determined using the crack crack opening area area of the CRDM annulus corresponding CADMannulus corresponding to aa radial gap of only 1 mil. The report also concludes that, that. should a CRDM nozzle CROM nozzle have have a through-wall through-wall crack, a leak rate of 0.04 gpm to less than 1 gpm will result crack,aleak in significant significant 'accumulation accumulation of boric acid crystals.

((rystals.

A.2 Axial Flaws Axial Raws Within the J-Groove Weld Flaws should grow axially through the J-groove J-groove weld due to the nature of the the stresses in the J-groove J-groove weld. ' Fora For a PWSCC-type PWSCC-type crack, it may break break the the surface surface as a very tight or pinhole-type pinhole-type crack crack in the annulus region.

region. These typestypes of cracks would result ina in a low leakage leakage as has, for example, been been observed during during the visual inspection CRDM nozzle number 21 at ONS-1 In December inspection of CROM December 2000 2000 (Reference 17) .and and at ON8-3 ONS-3 In February 2001. 2001. The maximum maximum amount of boric boric acid crystals crystals observed aroundaround the base of the ON8-1 ONS-1 CROMCRDM nozzlenozzle approximately 0.5 in3 ,, signifying a very low leakage rate through number 21 was approximately the crack. OnlyOnly small quantities of boricboric acid crystals were present present on the ONS- ONS-2, ONS-3, and ANO-1ANO-1 RV heads, as well.

Page 52 of 56 NRC004-1223 NRC004-1223

• • NON-PROPRIETARY

.... NON-PROPRIETARY - *- 51-5012567-01 51-5012567-01 RV Hea~ Nozzle and Weld Safety Assessment RV Head Nozzle and Weld Safety Assessment A.3 A.S Circumferential Flaw Extemal Circumferential External Flaw in CRDM Nozzle in CRDM Nozzje An assessment of An assessment external circumferential of external circumferential crack growth in crack growth in the CAOM nozzle the CRDM nozzle above the J-groove weld was addressed in Reference 7.11 it is postulated that above the J-groove weld was addressed in Reference 7. If it is postulated that aa circumferential crack circumferential through-wall and propagates through-wall crack propagates and grows clrcumferentially along gr9ws circumferentially along the weld-nozzle the interface region, weld-nozzle interface region, the the potential potential safety concern is detachment safetyconcemis detachment of of the the upper nozzle from upper nozzle from the lower nozzle the lower nozzle section section and and itsits ejection ejection from from thethe closure closure head. However, detection head. However, detection of of leakage leakage prior prior to to tube tube failure failure is predicted to is predicted to occur.

occur.

Based Based on on aa limit limit load analysisof load analysis of the CROM nozzle the CRDM geometry, the nozzle geometry, the net section limit net section limit ligament is less than 25%. Postulating that a large portion of the nozzle cross-ligament is less than 25%. Postulating that a large portion of the nozzle cross-contains aa through-wall section contains section circumferential crack, through-wall circumferential crack, there there is ample room is ample room for for leakage to occur before approaching the net section limit ligament. This will leakage to occur before approaching the net section limit ligament. This will sufficient leakage allow sufficient allow Jeakage of of steam through this steam through this large large crack crack to to bebe detectable, detectable, thereby thereby providing providing ample wamlng to ample warning to prevent prevent the the failure failure of of the nozzle. The the nozzle. The flow flow rates rates werewere predicted (without consideration pred!cted (without consideration of of potential "leak-plugging" in potentiaIDleak-plugglng" in aa narrow annulus) narrow annulus) for for aa six-inch six-inch circumferential through-wall crack circumferential through-wall (nearly 50%

crack (nearly 50% of of

.the circumferential extent, the circumferential extent, as as observed observed in in nozzle nozzle number number 56 56 at ONS-S). For at ONS-3). For annulus clearances annulus clearances of of 0.001 0.001 inch, 0.0016 Inch inch, 0.0016 Inch and 0.002 inch and 0.002 Inch (to cover cover the the ranges ranges of of the the predicted clearances during predicted clearances during normal normal steady steady statestate operation operation for for thethe center nozzle center nozzle to to the 9Utermost nozzle),

the outermost nozzle),the the leakage leakage rates rates were determined to were determined to bebe 0.4 0.4 gpm, 0.8 gpm and gpm,O.Sgpm and 1.2 gpm, respectively.

1.2 gpm, respectively.

A.4 A.4 Lack of Lack of Weld Weld Fusion Fusion AreasAreas In In the the J-Groove i:I-Groove Weld Weld The allowable The allowable lack lack of of weld fusion areas weld fusion areas In in the J-groove weld the J-groove weld of the B&WOG of the B&WOG plants was addressed in ReferenceS. Framatome ANP performed an "spection plants was addressed in Reference 8. Framatome ANP performed an 'spection of of the nozzle.,ta-vesse! head the nozzle-to-vessel head weldswelds In In aa section section of Midland Unit of Midland Unit 1, which Is 1, which is typical typical of the B&WOG plants. The of the B&WOG plants. The Inspections revealed that inspections revealed that the majority majority of of the the indications were indications located at were located at thethe CRDM nozzle-ta-weld Interface, CRDM nozzle-to-weld Interface, and and all all indications were indications were less less than than 22 inchesinches (51 (51 mm) long. long. Most Most of the indicationsindications detected detected In in the the Midland Midland welds welds are believed to be are believed be slag slag Inclusions, Inclusions, with with a.a fewer fewer number of areas indicating lack of fusion number of areas indicating lack of fusion of the weld zone. of the weld zone. The two two areas areas of concern for concern for the lack fusion are lack of fusion are the the CRDM nozzle-to-weld interface CROM nozzle-to-weld interface and .and the the head-ta-weldinterface. Both these areas were evaluated to determine the head-to-weld interface. Both these areas were evaluated to determine the minimum minimum weld weld area area required required to to meet meet the the ASME ASME Code Code primary primary shear shear stress stress limitslimits (i.e.,

(i.e., allowable allowable lack fusion size) and lack of fusion and to determine determine the the weld weld area area required required to limit limit the stress to shear stress the"shear to the shear shear flowflow stress stress (i.e.,(I.e~. critical critical lack lack of of fusion size). It was demonstrated that the was demonstrated the CRDM nozzle-ta-weld interface CAOM nozzle-to-weld interface was was more more limiting.

limiting.

The results showed The showed that approximately 67%

that approximately (corresponding to 8.4 67% (corresponding inches of S.4 inches circumferential extent) of circumferential of the total weld weld areaarea may be unfused unfused and still meet meet the the ASME Code shear stress limit Similarly, using ASME Code shear stress limit. SImilarly, using the Tresca shear flow stress the Tresca shear flow stress criteria, It was shown shown that that 85% (corresponding to 850/0 (corresponding 10.6 inches of circumferential to 10.6 extent) extent) of the total weld area may be weld area may be unfused and unfused and still have sufficient strength to have sufficient prevent aa catastrophic prevent catastrophic failure.

Page 53 of 56 NRC004-1224 NRC004-1224

• al

- W, NON-PROPRIETARY NON-PROPRIETARy ... *6**

• 51-5012567-01 51-5012567-01 RV Head Head Nozzle and Weld Safety Safety Assessment The allowable allowable lack of fusion size and indeed the critical lack of fusion size have have circumferential crack lengths significant circumferential lengths such that sufficient leakage can be sufficient leakage be demonstrated for these types of cracks.

A leakage leakage assessment for this postulated postulated circumferential circumferential crack in the weld was was performed using the methodology methodology very similar to that described described in Section A.1. A.1.

The only difference difference is that only one leakage path is considered considered which represents represents the annulus. The flow is assumed to choke at the exit of the annulus.

The crack lengths required to achieve achieve the leak detection capability rate of 19pm detection capability 1 gpm were determined determined for annulus clearances clearances of 0.0016 0.0016 inches and 0.002 inches.

These annulus clearances clearances correspond to the average radial gaps during steady state normal normal operating operating conditions for the center and outermost CRDM nozzles, respectively. ItIt was determined determined that a crack length of 7.5 inches in the center center nozzle (annulus of 0.0016 inch) 0.0016 inch) is required required to achieve achieve a 1 gpmleak rate.

gpm leak Similarly.

Similarly, it was determined determined that a crack length of 5.0 inches inches in outermost the outermost nozzle nozzle (annulus of 0.002 inch) is required to achieve achieve a 1 gpm leak rate.

~.

Page 54 of 56 NRC004-1225 NRC004-1225

• • • • NON-PROPRIETARY NON-PROPRIETARY **** **** 51-5012567-01 51-5012567.;01 RV RV Head Head NozzleNozzle andand Weld Weld Safety Safety Assessment Assessment Figure FigureA-1. A-1. Radial Radial Clearance Clearance for Center Center Nozzle Nozzle 0.0035 .------,...------r------.,.-------r--------r-------r-------,

0.003 - - - - -..*- _. '---' .-....... _....

0.003 .. .. .

0.0025 0.0025 1--,~---I__---__i----_+-------_+-------_I_------_4_----~

.5

§ 0.002 _.- ----_._ _ .. .......-....... _. --..- ........... _. _.. _........ --- -----...... ....- ._---_._....------

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IB 0.0015 0.0015 I------f-~~~-+~----t----- ----I-~------+~-~~--+---~

0.001 0.001 0.051 0.0005 ~--------_r--------~r-------;-------~--------+---------+----~--~

o 0 ~--------~~------~~------~--------~--------~------__~_________ J 0o 1 22 3 4 5 6 7 Distance from Bottom of Shrink Fit. In Shrink Fit, in Z

0 oo Page 55 of 56 o

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• • • • 51-5012567-01 51-5012567-01' RV Head Head Nozzle and Weld Safety Assessment

• ,.;.it Figure A-2. Clearance for Outermost A*2. Radial Clearance Outermost Nozzle Nozzle 0;0035 I..

0.003

~ v----

~

C 0.0025

~

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.5 8 ~

0.002 i(,)

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u 56 Page 56 of 56 o

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Docket Docket Number Number 50-346 50-346 License License Number Nurnber NPF-3 NPF-3 Serial NumbCr 2743 Serial Number 2743 Attachment 2 Page Page 1J

- Replacement Replacement Attachment Attachment 55 for for Letter Letter Serial Serial Number Number 2735 2735 Proprietary Version of Proprietary Version of Structural Integrity Associates, Structural IntegrityAssociates, Inc. Calculation, CaJculation, File File Number W-ENTP- I IQ-306, "Finite NumberW-ENTP-IIQ-306, "Finite Element Element Gap Gap Analysis Analysis of ofCRDM CRDM Penetrations Penetrations (Davis-Besse),"

(Davis-Besse),"

NRC004-1228 NRC004-1228

• tJ RAL STRUCTU STRUCTU CALCULATION CALCULATION W-ENTP-I IQ-306 FILE No: W-ENTP-llQ-306 INTEGRITY PACKAGE PROJECT No: W ~ENTP-ll Q W-ENTP-lIQ Associates, Inc.

PROJECT NAME: Davis-Besse Technical Response to NRC Bulletin 2001-01 2001-01 CLIENT: First Energy CALCULATION TITLE: Finite Element Gap Gap.Analysis of CRDM Penetrations (Davis-Besse)

Analysis ofCRDM PROBLEM STATEMF!'I'T PROBLEM STATEMF NT OR OBJECTIVE OBJECTIVE OF THE CALCULATION:

CALCULATION:

Develop a finite element element model of the top head and CRDM penetrations for Davis-Besse Nuclear Nuclear Power Station. The model is then used to evaluate the gaps between the CRDM tubes and the hemispherical normal operating conditions.

hemispherical head during normaJoperating Project Preparer(s)

Preparer(s) &&

J:.

Document Document Affected Affeded Revision Description Re\'ision Description Mgr. Checker(s)

Checker:(s)

Revision Pages Approval Approval Signatures &&

Signature && Date Date 0 1-24 1-24 Original Issue Original Al -A3 Al-A3 B1 -B2 BI-BZ Cl-C6 CI-C6 DI -D3 Dl-D3 Project CD-Rom CD-Rom

.. PAGE 1 of 24 PAGE--L-of~

FramatomeANP Framatome ANP Proprietary Proprietary NRC004-1229 NRC004-1229

1.0 Problem 1.0 Problem Develop aa fmite Develop finite element element model model of ofthe the top top head head and and CRDM CRDM penetrations penetrations for forDavis-Besse Davis-Besse Nuclear Nuclear Power Station. The Power Station. The model model is is then then used used to to evaluate evaluate thethe gaps gaps betw.een between the the CRDM CRDM tubes tubes and and the the hemispherical head hemispherical head during during nonnal normal operating operating conditions.

conditions.

2.0 Finite 2.0 Finite Element Model Model A finite A finite element element model model hashas been been constructed constructed using using the the ANSYS ANSYS finite element software finite element software package package [1].

[1].

The model The model includes the the upper hemispherical hemispherical bead, head, the upper closure flange and closure flange and the the CRDM housing housing tubes. Due Due to to the the symmetrical symmetrical nature nature of upper headhead structure structure and the layout and the layout of ofthe the CRDM tubes.tubes, only 45° only 450 of of the the total total circwnference circumference was was modeled.

modeled. Additional details are are described described inin the the following following resulting model can be seen sections. The resulting seen in in Figure 1. 1.

2.1 Hemispherical 2.1 Hemispherical HeadfUpper Head/Upper Closure Flange Flange References 2 and 33 provided the the closure closure flange ani.nemisphericalhead an," nemispherical head dimensions used in the finite element element model. The flange flange and and head head were c,nstructed c..,nstructed using the ANSYS 8*node 8-node SOLID45 SOLED45 elements. The The following assumptions were made during the construction of the segment of the finite element model:

• " The The clad clad material was included as base metal for determination of dimensions and modeling.

" The clad

• The clad was assumed to be aa constant 3/ 16 inches thick throughout the structure.

3/16

• The bottom

• The bottom face face of the closure flange does not specifically model the contact surface.

The bottom face remains plain and the contact cobtact surface will be simulated via gap elements elements (described later (described in the loads section of this calculation).

Jaterin calculation).

• " The hemispherical hemispherical head to closure flange fillet radius on the outside surface was assumed 1 assumed to be 6614 /4 inches. .

• " The closure closure bolt holes were not specifically specifically modeled, thus the closure closure flange is a solid solid structure. However, structure. However, the the locations of the bolt holes (there were 7-1/2 holes (there were 7-112 holes in the 450 45° segment segment modeled) modeled) were were modeled modeled to provide provide loading loading points for the bolt bolt preload.

preload.

• • . Some additional assumptions/variations assumptionslvariations in the hemispherical hemispherical head head will be be described in in the following following section section of the CRDM housings.

See See Figure Figure 2 for thethe dimensions dimensions used.for used Jor the the hemispherical hemispherical head head and closure flange.

Revision o PreparerlDate RLB 1O~8*01 CheckerlDate STC lO'()8'()1 FiJe No. W~ENTP-llQ-306 Framatome FramatomeANP ANPProprietary Proprietary NRC004-1230 NRC004-1230

2.2 CRDM Housing Housing Tube Penetrations Penetrations A total of 13 CRDM housing tube penetrations where modeled. They also were modeled modeled using the ANSYS 8-nodeSOLID45 element. Based on the 45° 8-node SOLID45 element. section modeled, the following tube 450 section configurations were actually included:

configurations incJuded:

• • 11 is is modeled modeled as 45° 450 (top dead center) (Tube 1)

• e 7 are modeled as 180 I g0 0 (along the symmetry 3,6,11,22,27,47,58) symmetry boundary) (Tubes 3,6,11,22,27,47,58)

• • 5 are fully modeled as 3600 fullymode1ed 3600 (Tubes 15,31,39,51,63) 15,31,39,51,63)

See Figure 3 for the locations of penetrations modeled. The dimensions were provided in the penetrations ofthe in Reference 3.

Reference 2.2.1 Hemispherical Hemispherical Head CR:JM CR:M Penetration Penetration Dimensions Dimensions Based on Reference Reff'rence 3, the thepenetlation penetiation hole in the hemispherical head is 4.0 inches in diameter. The resulting resultinginterfer interfer ;nee nce fit region begins at the .outside outside surface of the extend.; down to the toe ofthe hemispherical head and extend&

hemispherical I-groove weld (at the top of the weld of the J-groove butter - see Figure 4a).

For this analysis. region was modeled interference fit region analysis, the interference modeled as beginning at the top edge of the weld butter CRDM to hemispherical hemispherical head weld and extending extending to the outer surface of the head. 'The The weld was not specifically specifically modeled. This Tesulted resulted in layout as shown in Figure 4b.

2.2.2 CRDM Tube DimensionsDimensions Reference 4, the CRDM tube outside diameter Per Reference diameter at the penetration is is 4.025 inches with anan inner diameter at 2.765 inches.

For this analysis, the tube outside diameter is set at a constant 3.998 inches and inner analysis,tbe inner diameter maintain the original 0.63 inch wall diameter at 2.738 inches (to maintain wall thickness).

thickness). The outside CRDM CRDM diameter of 3.998 inches allows for a 0.001 inch radial gap between diameter of3.998 between the CRDM CRDM tube and the hemispherical hemispherical head hole (modeled (modeled at 4.00 inches diameter). This gap was necessarynecessary support CONTAC52 to support CONTAC52 elements, which which were used to simulate simulate the interference between interference fit between the CRDM and the hemispherical hemispherical head penetration holes (see penetration boles (see Section 5.6 for additional details on the interference interference loading).

The effects of the modeled reduction in outside diameter diameter (and corresponding corresponding inside were-considered diameter) were considered insignificant insignificant for this gap evaluation. In addition, due to the variation in interference interference values (Reference S) 5) between between each tube, any variations variations from the References 3 and 4 were further minimized' drawing dimensions from References mninimized.

Revision o PreparerlDate RLB RIB l()"o8"()l 10-08-01 CheckerlDate STC 1 ()'oS-Ol File No. W-ENTP-IIQ~306 Page 3 of 24 Framatome Framatome ANP Proprietary ANPProprietary NRCO04-1231 NRC004-1231

In addition, neither the tube expansion nor the bolted flange COIUlectiOns connections (both (both of which which occur outside the reactor vessel) occur vessel) are modeled and the total total height of the tube beyond the hemispherical head is arbitrary.

Finally, the actual CRDM tubes are intended to project through and slightly into the Finally, the actual CRDM tubes are intended to project through and slightly into the hemispherical head. 1his This projection was not modeled.

2.2-3 CRDM-to-Head'Veld 2.2.3 CRDM-to-Head Weld A key set set of dimensions was the varying height of the CRDM tube to hemispherical head weld. The we1d. The weld weld height height varies around the circumference of the tube based on each tube's tube's position position on on the the hemispherical hemispherical head (see Figure 4a and Reference Reference 3).

While the specific weld and weld material While material were not modeled, the weld attacbment attachment height is important important and must be included. To determine these heights, the c-bservations and following c*bs.ervations assumptions were made: .

• 9 The 3/ 16 inch butter (thou*.,h 3/16 (thou ,h not modeled modeled specifically) was not considered considered as part of of the weld attachment height calculation. The butter was not modeled but considered considered part of the hemispherical hemispherical head.

• 9 The vertical distance from the top edge of the weld (bottom of the butter) to the verticaJ distance comer corner of the base metal of the hemispherical head is a constant constant 25/ 32 inches anywhere 25/32 on the weld and for every tube (Reference (Reference 3).

• • TheThe diameter diameter between the opposite centers of curvature for for all of the CRDM tube to hemispherical 1 hemispherical head welds is a constant 441/4/ 4 inches.

• & The weld height varies linearly around-the aroundlhe circumference circumference [3].

Based Based on Reference 6, the radial extent of ofthe J-grooveweld J-groove weld (including(including the butter) from the radius of radius curvaturecurvature to the outside surface outside surface of the base metal is 0.8606 inches. The resulting .

edge-to-edge diameter of the weld prep is therefore 4.25 inches + + 2 ** 0.8606 0.8606.inches inches or or 5.9712 inches.

A simple ANSYS model was thus developed A developed that penetrates penetrates the inner surface surface of oftbe the hemispherical head hemispherical bead at eacb at each tube location using the diameter of usingtbe 5.9712 inches. For this model ofS.9712 it it is necessary to is necessary to specifically specifically exclude exclude the inner clad, resulting in an inner inner surface surface radius forfor the hemispherical head of 87 1 inches [2].

871/4 hemispherical bead /. inches [2]. The resulting intersections intersections of ofthe the 5.9712 inch inch diameter diameter penetration penetration and the inner hemispherical hemispherica1 head surface were were then shifted 25/ inches 25/32 inches up to determine detennine the top height of the welds. The ANSYS file used for this study is named named WELDINP WELD.INP and included included with the Project Project CD-ROM.

CD*ROM. The resulting J'eSlllting heights are also included Appendix A.

in Appendix A.

Revision 0 tJ VPreparer/Date PreparerlDate Checker/Date CheckerlDate File No. W-ENTP-1IQ-306

_FileNo.

RLBIO-08-01 RILB 10-08-01 STC 10-,8-01.

STC W-ENTP-IIQ-306 10-08-01 .' .'

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final finite element model the height values In the final values detemrined determined above above were .used used in conjunction with the actualactual modeled CRDM tubes to create aa series of ofangled angled planes, which were used to dividedivide the CRDM tube at at the top of the weld (see (see Figure 4b).

The final final weld weld connection between the hemispherical .bead head and the CRDM tubes is via a series of degree-of-freedom degree-of-freedom couples between the nodes along the inner surface of the bole hole in hemispherical head and the outer smface the bemispherical surface nodes of the CRDM tubes. These couples are only applied along the modeled weld height and can be seen seen in Figure s.5.

3.0 Materials Reference 2 indicates the following materials were used for the modeled components. Note that the closure stud was not actually modeled and its properties are not included in this evaluation.

Component J Material Upper-Head Upper Head SA-S"-\3 SA-543 Grade B Class 11 . ".-

Closure Flange ~A-S08Class SA-508 Class 2 CRDM Housing Tube SB-167 SB.;167 (Alloy 600)

(A1loy60Q) 1 Closure Stud I SA-540 SA-S40 B23 Class 3 B23Class No welds were specifically specifically modeled nor were the weld materials included. Where material changes changes occur across welds, the material was simplyrnodeled simply modeled as an instant change. In the case of the hemispherical head to closure flange, the material is assumed to change at the end of the fiUet fillet region farthest from the closure flange.

The material material properties properties used for this evaluation evaluation are based on the 1989 ASME Code [7] for a temperature of 600*F.

600°F. The properties properties used are indicated indicated in the following table:

Modulus Modulus of of Elasticity Elasticity Mean Coefficient of Thermal Material ,_E, E, psi Expansion, . 1 in/m/*F ExpaDsion.cx. infmfOF SA-533 Grade B Class I1 SA-S33 26.4e6 7.83e-6 SA-508 SA-SOS Class 2 26.4e6 7.42e-6 SB-I 67 (Alloy 600)

SB-167 28.7e6 7.82e-6 7.82e-6 A Poisson's Poisson's Ratio of 0.3 was usedJor used for all aU of the materials, as was the metal materials, as. metal density of 0.283 lb/in3 .

densityofO.283Ib/in'.

4.0 Mechanical Mechanical Boundary Boundary Conditions Revision Revision } ()

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Preparer/Date PreparerlDate CheckerfD*

CheckerlD*

, File RB STC File No. W-ENTP-1IQ-306 W -ENTP-IIQ-306 10-08-01" RLBIO-oS-ol STC 10-os-01 10-08-01 .- "~

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Symmetry boundary Symmetry conditions were boundary conditions were applied along the applied along circumferential free the circumferential free ends ends of of the model. See the model. See Figure 6 for these boundaries.

Figure 6 for these boundaries.

In In the case of the case of the the closure closure surface, surface, aa series series of CONTAC52 gap ofCONTAC52 elements were gap elements were developed developed at the at the compressionsurface (see Figure 2). The gap elements attach the compression compression surface (see Figure 2). The gap elements attach the compression surface of the closure surface of the closure flange flange toto aa series series of of nodes nodes that that are are fixed fixed in in the the vertical vertical direction.

direction. These These nodes nodes simulate simulate thethe compression compression surface surface of of the the lower lower (un-modeled)

(un-modeled) flange. flange. The The twotwo nodes nodes that that make make up this contact upthls contact region are 0.1 inches below the compression surface of themodeled legion are 0.1 inches below the compression surface of the modeled flange, but behave as if flange, but behave as ifthey they are are in direct contact. In addition, the node pairs of each gap element in direct contact. In addition, the node pairs of each gap element are coupled for horizontal are coupled for horizontal translations translations and and have have aa weak weak spring spring element element (COMBINI4, (COMBIN14, k=100 k=100 lb/in)

Ib/in) between between them.

them. These These additions are included to provide initial additions are included to provide initial numeric numeric stability stability inin the the analysis.

analysis. See See Figure Figure 77 for for applied applied couples and vertical restraint on gap couples and vertical restraint on gap elements. elements.

5.0 5.0 Loading Loading Two Two gapgap uvaluations I,;valuationswere were pp ,rfomed rformed underunder normal normal operating operating conditions.

conditions. TheThe only only variation variation between the two evaluation.; was the interference loads between the two evaluation; was the interference loads between the CRDM tube and between the CRDM tube and the the hemispherical hemispherical head. head. The The loads loads that that exist exist for for the the normal nonna1 operating operating condition condition areare defined defined in in the the following sections as are interference following sections as are interference loads used in loads used in the the two two gap gap evaluations.

evaluations. .

5.1 5~1 Temperature Temperature AA uniform temperature of unifomtemperature of605°F 605 DF [8)

[8] was was applied applied over over the the entire entire model model with with the the stress stress free free temperature being temperature being 70°F. 70°F.

5.2 5.2 Pressure Pressure The The normal nonnaI operating operating pressure pressmc isis 2155 2155 psipsi forfor Davis-Besse, Davis-Besse,Unit Unit 11[8].

[8]. The pressure was Theprcssure was applied applied to the inside surface of the hemispherical head, the hemispherical to the inside surface ofthe hemispherica1 head. the hemispherical head side end of the CRDM head side end of the CRDM tube, tube, and and to to the theflange flange closure closurefaceface out out to to aa radius radiusof 84.8115 of84.81 15 inches inches [2].

[2].

In In addition, addition, aa cap cappressure pressurewas was applied applied to to the the outside outside free free end end ofofthe theCRDM CRDMtubes tubes to to simulate simulate line load in each tube. The pressure was line load in each tube. The pressure was calculated as: calculated as:

P-r.n2 2155.1.372 (rMdt _- r-,, 2 (2.02 -. 1.372) =1905.1 psi Note that the applied cap load was actually applied in the negative direction in ANSYS, thus Note that the applied cap load was actually applied in the negativcdirection in ANSYS,thus providing providingaatraction traction load.

load. See SeeFigure Figure88for forthe the applied applied pressure pressuresurfaces.

surfaces.

5.3 S.3 Closure ClosureBolt BoltLoad Load Revision0 Revision o ..

P PreparerlDate RLB10-08-01 Checker/Date Preparer/Date STC 10..()8-O1 RLB 10-08-01

~ ,rCh~~~e-rlD~m~e~-*S-JC-.~1-0-..()~g-O~1~--------~----~--~--------------~

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A total closure bolt load of84.0e6Ibs of 84.0e6 lbs is specified specified in Reference Reference 8. As there are Ita total of60of 60 bolts, the total load per bolt is 1.4e6lbs.

1.4e6 lbs. The closure bolt load was simulatedsimulated by applying a pressure pressure load at the top of the flange on each of the bolt hole locations that were modeled. The area of each each hole is 7t-3.S 7t*3.512 or 38.48 38.48in in2; applied pressure is 36378.27 psi. See Figure 9 2; thus the applied for the applied pressures pressures for for the bolt load simulation.

simulation Using pressure allows a more rapid model development and should involve involve no significant loss of accuracy accuracy since the areasofintetest areas of interest were the CRDM tubes. In addition,addition. the overall overall size of the closure flange relative to the rest of the head head minimizes any stiffness changes anystifIhess changes had the holes been modeled and the actual studs included.

5.4 Gasket/Spring Gasket/Spring Loads During closure, there are three three other activeloads active Joads applied to the upper flange; two gasket loads and a spring load. The gasket squash squash loads and theirradiusofat:)lication their radius of aplication were defineddefined in References 8 and 2, respectively, as:

References .

• " Inner Gasket: Squash Load =

Gasket: Total Squasb = 400,000 lb. at,at Lradius of84.8115 of 84.8115 inches

• " Outer Gasket: Total Squash Squash Load = 407,600 lb. a~ a: a radius ofof86.4365 86.4365 incbes inches gasket loads are applied The gasket applied as aa series ofnodal loads at the bottom of the flange in a positive positive vertical vertical direction. The total squash load for a 45° 450 section section of the model is 50,000 50,000 lbs Ibsand 50,950 and 50,950 lbs for the inner iIUler and outer gaskets.

gaskets, respectiveiy.

respectively. At each radius of load application there are 37 37 equally spaced circumferential circumferential nodes (for a total of74 of 74 nodes in two lines, 37 for the inner gasketgasket and 37 for the outer gasket with the two rows of nodes lying side by side in the finite element model). o d e l ) . , . ,.

m The total load on the inner 3S 35 nodes for each of the gaskets was 1,388.89lbs 1,388.89 lbs for the inner and 1,415.28 1,41 5.28lbs for tlieouter lbs for the outer gasket. The 2 symmetry symmetry edge nodes for the inner gasket are loaded loaded with 694.45 Ibs lbs while the outer gasket symmetry symmetry nodes received 707.64 707.64 lbs.

Ibs. Figure 10 shows the applied load as two sets sets of small upward arrowsmows nearest the outside edge oftbe of the closure flange.

The spring load (also referred Tbespring referred to as "ledge" load) load) is the reaction of the plenum oftbe plenum cover cover and core support shield assembly assembly (not modeled for this evaluation) to the applied closure stud preload.

closure.stud For the the.upper upper closure flange the response load and its radius of application was defined in Reference 8 as:

Reference

• • Spring Load: Total Load Load:;;= 6.0e6 6.0e61b.lb. at an assumed radius of80.Sof 80.5 inches The spring load was simulated simulated in the same mannermanner-as as the gasket loadings; it was simulated simulated with a series of evenly spaced circumferential circumferential nodal loads. The total load for the 450 45° model was 750,000 lbs. The inner 35 nodes therefore received a load of20,833.33 750.000 of20,833.33 Ibs lbs in the positive vertical direction while the symmetry edge nodes were loaded at I10,416.67 lbs. Figure 10 shows 0,416.671bs.

Revision 0o

• • Preparer/Date PreparerlDate RLB 10-0-1 lO"()8-Oi Checker/Date CheckerlDate STC 10-08-01 STC 10-0S"()l fileNo. W-ENTP~IIQ-306 File.No. W-ENTP,-1IQ-306 Page 7 of 24 Framatome.ANP Proprietary FramatomeANP Pr()prietary NRC004-1235 NRC004-1235

the applied load as a row of large upward arrows furthest from the outside edge of the closure the applied load as a row oflarge upward arrows furthest from the outSide edge of the closure flaDge.

flange.

5.5 5.5 Deadweight Deadweight Load Load 1-g positive vertical acceleration A l-g acceleration (in ANSYS a positive acceleration produces the desired acceleration produces deadweight load) was applied to the model to simulate simulate gravity for the deadweight .Ioad.load.

Although the entire tubes are not modeled nor are any other extraneous components, the total components, effect is expected to be minimal in comparison comparison to the total weight of the head.

5.6 5..6 CRDl\*1 Interference Load CRDM Housing Interference Load The final load applied to the finite fmite element element model was the interference loads between between the tube surfarms and the inside interference outs;desurfar-:s outside interference zone of the hemispherical hemispherical bead. was head. This load was the only load th~ cbangedbetween th;". changed between the two gap analyses. The interference interference dimensions dimensions at the top and 0..- each the bottom ol-each tube -.

t Davis-Besse, "t Davis-Besse. Unit 1 was proVided Reference 5.

provided in Reference The interf !rence values for each tube will be varied rence values varied linearly linearly from the top edge of the bemispherical hemispherical head down d"vm to the bottom of the interference zone Gust (just above the weld). Note that because of the layout of the interference interference zone the interference interference values will vary around the circumference circumference of the tube for a given height. This is due to the changing changing interference interference zone height of each tube as a result of the slope of the hemispherical hemispherical head (see Figure 4)

The first analysis will be used .to frrst gap analysiswiU to support a leak rate calculation based on the gap.openings gap.openings between the CRDM and the hemispherical As such the worst (or largest) interference hemispherical head. }psuch interference modeled to minimize values were modeled opening and thus leak rate. In addition, with only 13 of minimize gap operung of the 69 tubes modeled, the worst interference interference load for the corresponding tube sets was used.

Worst case for this analysis was the tube that had the greatest greatest top or bottom interference interference dimension.

dimension.

I lists the modeled tube numbers, the tube numbers in the corresponding Table l1ists corresponding tube set, the worst case tube, interference dimensions for that tube..All tube. and the resulting interference tube. AIl of the tube interference values are included interference Appendix B. The final ANSYS input file for the gap included in Appendix evaluation was named DBCRDMINP and is included on the project CD-Rom.

named DBCRDMlNP Table Table I1 Revision o PreparerlDate RLB 10~8-O1 CheckerlDati. STC.lO-O8-O I ~ ... -

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Interference Values Interference Values for for Gap Gap Evaluation Supporting Supporting Leak Leak Rate Evaluation Evaluation Modeled Corresponding Worst Diametrical Interference Diametrical Interference Modeled Corresponding Worst (Gan) Dimensions (in)'

Tube ##

Tube__________________Top Tube Tube ## Tube Tube ## TOP e otsomn Bottom*

Top Bottom 11 33 2,4,5 2.4.5

-10.0001 1 22 0.0001 0.0019 0.0019 0.0012 0.0012 0.0020 0.0020 66 7,8,9 7.89 66 0.0012 0.0012 0.0006 0.0006 11 11 10,12,13 10,12,13 10 10 0.0013 0.0013 0.0002 0.0002 15 15 14,16,17,18,19,20,21 14,16,17,18,19,20.21 17 17 0.0014 0.0014 0.0013 0.0013 22 22 23,24,25 23,24.25 24 24 0.0016 0.0016 0.0004 0.0004CGap) (Gap) 27 27 26,28,29 26.28.29 27 27 0.0011 0.0011 0.0005 O.OOOS 31 31 30,32,33,34,35,36,37 30,32,33,34.35,36,37 33 33 0.0018 0.0018 0.0003 0.0003 39 39 38,40,41,42,43,44,45 38,40.41.42.43.44.45 45 45 0.0014 0.0014 6.0011 G.OOI1 47 47 46,48,49 46,48,49 48 48 0.0009 0.0009 ,..0013

\.. 0013 51 51 50,52,53,54,55,56,57 50.52.53,54.55.56,57 50 SO 0.0021 0.0021 0.0010 0.0010 58 58 59,60,61 59.60.61 61 61 0.0012 0.0012 0.0003 0.0003 63 63 62,64,65,66,67,68,69 62,64,65,66.67,68,69 63 63 0.0014 0.0014 0.0015 0.0015 The The second second gap gap analysis analysis will will be be used used toto support supportaa fracture fracture mechanics mechanics evaluation evaluation of ofthe the CRDM CRDM weld cracking.

weld cracking. To To support support this this evaluation evaluation thethe least Jeast (or(or smallest) smallest) interference interference values values were were modeled modeled to to maximize maximize gap gap opening opening and and thus thus maximize maximize crack crack opening openingin in the the follow-on follow-on flaw flaw evaluation. Again, with only 13 of the 69 tubes modeled, the least interference evaluation. Again, with only 13 of the 69 tubes modeled, the least interference load forthe load for the corresponding corresponding tube tube sets sets was was used.

used. Least Least for for this this analysis analysis waswas the thetube tube that thathad hadthethe smallest smallest top top or or bottom bottom interference interferencedimension. Note that in dimen!rion.Notethat in aa number numberof ofcases cases there therewas wasno no interference interference but but an an actual gap instead.

actuaIgap instead. Any Anyactual actual gaps gaps were were included included in in the themodel.

model.

Table Table22 lists liststhemodeled the modeledtube tubenumbers, nwnbers, the the tube tubenumbers numbers in inthe the corresponding correspondingtube tubeset, set,the thebest best .'

case tube, and the resulting interference dimensions (or case tube, and the resulting interference dimensions (or gap dimensions) forgap dimensions) forthat thattube.

tube.All Allofofthe the tube tube interference interferencevaluesvaluesare are included included in inAppendix AppendixB. The final ANSYS B.Thefinal ANSYSinputinputfile file for forthe thegap gap evaluation evaluation was wasnamed named DBCRDM-O.INP DBCRDM-O.1NPand and isisincluded includedon on the theproject projectCD-Rom.

CD-~m.

Table Table22 Revision 0 tJ PreparerlDa~e CheckerlDate FileNo. W-ENTP-I1Q-306 RLBIO-08-01 STC 10-08-01 P82e 9 of 24 Framatome FramalomeANP ANPProprietary Proprietary NRC004-1237 NRC004-1237

Interference Values for Gap Evaluation Interference Supporting Flaw Evaluation Evaluation Supporting DiametricallDterference Diametrical Interference Modeled Modeled Corresponding Corresponding Best Best (Gap) Dimensions Dimensions (in)

(in)

Tube # Tube # Tube#

Tube # Top Bottom Top Bottom 1I 33

--_1 2,4,5 1.

55 0.0001 0.0001 0.0007 0.0007 0.0012 0.0012 0.0009 6 7,8,9 1,8,9 7 0.0001 0.0001 0.0009 11 10,12,13 10,12,13 13 0.0009 0.0001 Gap 15 14,16,17,18,19,20,21 14.16,17.18,19,20,21 14 0.0004 O.00Q4GapGap 0.0005 Gap 22 23,24,25 23.24,25 24 0.0016 0.0004 Gap 27 26,28,29 28 0.0004 0.0008 31 30,32~33,34,3S,36,37 30,32,33,34,35,36,37 35 0.0002 0.0010 Gap 0.0010 39 38,40.41.42,43,44,45 38,40,41,42,43,44,45 44 0.0012 0.0002 0.0002 47 46,48,49 49 .. 0.0002 0.0002 Gap 51 50,52,53,54.55,56,57 50.52,53,54.55,56,57 54 0.0000 0.0001 Gap 58 59,60,61 59.60,61 59 0.0008 0.0001 63 62,64.65,66,61,68,69 62,64,65,66,67,68,69 67*

67* 0.0005 0.0006

• Tube 61

• Tube 67 was was selected selected in in lieu lieu of Tube 65 of Tube 65 (Top (Top Interference Interference = = 0.0010 0.0010 inches, inches, Bottom Bottom Interference Interference = 0.0004), as the average interference of Tube 67 from top to bottom was lower than Tube 65.

For both evaluations, the application of the interference Forbotb interference (or gap) was via a CONTAC52 CONTAC52 gap element The CONTAC52 CONTAC52 element allows the entry of a negative gap value, which which is treated as an interference value rather than the typical typical positive physical gap. Each tube was thus modeled modeled with a series of gap elements simulating elements simuJatingthe the specific interference interference value. The interference interference values entered entered into ANSYS were halved, as the ANSYS element was established as a radial gap. The values were also evenly spaced down the interference zone and varied linearly from the top value to the bottom, resulting in a total of 5 sets of interference interference value interference interference values around the circumference of the tube circumference lube for each modeled modeled tube. The use of CONTAC52 elements in this ofCONTACS2 application application was verified in a separate study shown in Appendix C.

6.0 Results evaluation, a series of gap results were determined For each evaluation. detennined via a post-processing post-processing file named POST.

The post-processing post-processing file captured captured the element number, the gap opening and the gap's position element nwnber, relativ.e to its speci.fi~.tpbe.

relative specifip tube. Specific results for each evaluation evaluation are detailed in thefollowmg the following sections.

6.1 Gap Oening Opening Evaluation - Leakage Leakage Evaluation Revision Revision 0 VV Preparer/Date PreparerlDate Checker/Date CheckerlDate RLB 10-0"-0i RLB 10-08-01 STC 10-08-01 STC 10-08-01 File No. W-ENTP-11Q-306 l_FileNo.W-ENTP-IIQ-306 Page 10 of 24 FramatomeANP Framatome ANP Proprietary Proprietary NRC004-1238 NRC004-1238

F()r this evaluation, For evaluati()n, where where the the greatest greatest CRDM CRDM interference values were used, it was interference values was determined determined not have a vertical path that would that tubes I1 and 3 do not allow leakage would allow Jeakage (see Figure 11). In both (see Figure both cases, the blockage cases, blockage occurs just above the interference zone (i.e. just occurs at the bottom of the interference the weld). All AU of ofthe vertical path where the other tubes have a vertical leakage is possible.

where leakage Table 3 lists possible. Table each specific lists each specific tube's smallest gap opening tube's smallest along aa vertical path anywhere in the interference opening along zone whose gaps interference zone gaps are interference value is listed minimum interference are all open. For tubes 1 and 3 the minimum form of listed in the fonn ofa negative negative value.

value. A complete list A complete of tube list of tube results can can be be found in the Excel spreadsheet DB-GAP.XLS Excel spreadsheet (included (included on the project CD-Rom).

Table 3 Table Minimum Minimum Gap Results for Leakage Evaluation Results EvaJuatioD Tube Number TubeNumber Minimum Gap (inches)

Minimum (Intbes) 1 -0.00000367 (lnteTh.~ce)

-0.00000367 (Interference) 3 ~0.OOOO2483 (Interference)

-0.00002483 (Interference) 6 0.000073863 0.000073863 11 0.000012591 15 15 0.000011731 0.000011731 22 0.000019860 0.000019860 27 0.000081417 0.000081417 31

.31 0.000066524 0.000066524 39 0.000020384 0.000020384 47 0.000082758 0.000082758 51 " 0.000000682 0.000000682

.58

.58 0.000102970 0.000102970 63 0.000001171 As a result ofTube3's of Tube 3's lack ofa of a gap condition, further investigation investigation of the other tubes on the group of Tubes 2, 3,43, 4 and 5 were evaluated. The results of these evaluations are included in Appendix D of this calculation package.

Appendix 6.2 Gap Opening Evaluation -. - Fracture Mechanics Evaluation Mechanics Evaluation For this evaluation, where the least CRJ)MCRDM interference interference values were used (in (in some cases actual physical pbysical gaps were used), a series of gap values were determined future fracture determined to support futme mechanics evaluations. Specifically.

mechanics Specifically, gap infotmation interference zone information axially along the tube's interference infonnation axially along the tube's at the highest point of the weld (uphill side) and gap information tube's interference zone at the lowest POint interference (downhill side) were determined.

point of the weld (downhill detennined.

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Table 4 lists Iistseach tube's gap opening as it varies along a vertical path in the interference each specific tube?s interference zone at the uphill and downhill sideS. sides. A complete list of tube results can can be found in the Excel DB-BIG.XS (included on the project CD-Rom).

spreadsheet, DB*BIG.xLS spreadsheet, Table Table 4 Tube Gap Results for FractureFracture Mechanics Evaluation Evaluation CaD Gap (inches) at Interferene (inches)at Interference Zone Level Tube H11 Hill ToplMid TopIMid Top/Mid ddle MldIBot M ot MidIBot Midot Bottom

'# Side Top Top/Mid MIddle

1. 1

- #1 til #2 t#2 III 112

1. 2

.33 1 -

Uphill I 0.0017027 0.0024574 0.0024574 0.0016565 0.0023365 0.0023365 0.0015937 0.0015937 0.0021809 0.0021809 0.0015723 0;0020339 0.0020339 0.0015353 0.0015353 0.0018463 0.0018463 0.0011360 -0.0000020 0.0011360 0.0013500 0.000)556 0.0013500 0.0001556 Downill Do\\'Ilhi.ll 0.0003352 0.0005402 0.0007489 0.0007489 0.0010388 0.0010388 0.0013316 0.0011618 0.0001205 6 Uphill 0.0028988 0.0028988 0.0026593 0.0023999 0.0023999 0.0021418 0.0021418 0.0018442 0.0000598 0.0012589 0.0000598 Downhill -0.0000026

-0.0000026 0.0001903 0.0004334 0.0004334 0.0007807 0.0011477

.0.0011477 0.0010052 0.0000002 0.0010052 0.0000002 11 Uphill 0.0014143 0.0015783 0.0017222 0.0018428 0.0018428 0.0018881 0.0018881 0.(1016480 0.0016480 0.0006723 0.0006723 Downhill DoW1lhill 0.0008567 0.0010146 0.0010146 0.0012106 0.0012106 0.0015580 0.0019643 0.0019643 0,0006789 0.0017880 0.0006789 15 Uphill 0.0036170 0.0034250 0.0031924 0.0031924 0.0029331 0.0026183 0.0020375 0.0008296 0.0008296 DomV.l Dov.'!Ihill -0.0000355

-O.OOO035S 0.0001316 0.0005409 0.0005409 0.0011180 0.0017579 0.0017579 0.001756.1 0.0007555 0.0017561 0.0007555 22 Uphill 0.0019850 0.0019850 0.0021062 0.0021694 0.0021694 0.0021753 0.0021219 0.0018306 0.()008272 0.0018306 0.0008272 Downhill Dov.'Jl.b.iU -0.0000361 -0.0000139

-0.0000139 0.0002112 0.0002112 0.0008535 0.0017436 0.0017436 0.0019320 0.0008897 0.0008897 27 Uphill lJpbill 0.0016772 0.0017365 0.0017705 0.0017705 0.0017703 0.0016893 0.0016893 0.0013648 0.0002833 0.0002833 Downhill Dov.'Dhilt 0.0009714 0.0009983 0.0009983 0.0011108 0.0011108 0.0014537 0.0019235 0.0016472 0.0016472 0.0002996 0.0002996 31 Uphill 0.0033857 0.0033432 0.0033432 0.0032294 0.0032294 0.0030558 0.0028078 0.0028078 0.0023198 0.0011671 0.0023198 Downhull DownhIll -0.0000648 -0.0000090

-0.0000090 0.0004283 0.0011914 0.0020943 0.0021780 0.0011045 0.0021780 0,0011045 39 Uphill

.Uphill 0.0025787 0.0025887 0.0025887 0.0025279 0;0025279 0.0023936 0.0021773 0.0021713 0.0017300 0.0006024 0.0017300 0.0006024 Downhill -0.0000846

-0.0000&46 .-0.0000535

-0.0000535 0.0000485 0.0007372 0.0016514 0.0016514 0.0016877 0.0005377 0.0016877 47 Uphill 0.0018693 0.0020580 0.0020580 0.0022058 0.0022682 0.0022081 0.0022081 0.0019444 . 0.0008877 0.0008877 Downhill Dov.'Dhll1 0.0011666 0.0011666 0.0011483 0.0011483 0.0013298&

0.0013298'* 0.0018660 0.0018660 0.0025989 0.0023482 0.0023482 0.0009046 0.00()9046 51 Uphill 0.0039365 0.0038082 0.0038082 0.0018660 0.0018660 0.0032373 0.0028077 0.0028077 0.0021742 0.0008409 0.0021742 Downhill Downhm -0.0000887 -0.0000450

• -0.0000450 0.0002246 0.0002246 0.0010571 0.0020923 0.0020923 0.0020855 0.0007673 0.0007673 58 Uphill 0.0031386 0.0030919 0.0029563 0.0027254 0.0024084 0.0007196 0.0019036 0.0007196 Downhill DOWD.hilJ' -0.0000573 -0.0000342 0.0000799 0.0009045 0.0019880 0.0020242 0;0020242 0.0007205 63 Uphill 0.0034363 0.0034363 0.0033127 0.0030870 0.0027611 0.0027611 0.0023475 0.0017659 0.0017659 0.0005004 Downhill Downhm -0.0001121 .0.0000789

.0.0000789 -0.0000041 0.0008004 0.0008004 0.0018993 0.0018984 0.0018984 0.0004711 Revision0 Revision 0

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7.0 7.0 References References I) ANSYS Mechanical,

1) ANSYS Revision 5.7, Mechanical,Revision ANSYS Inc.,

S. ',ANSYS December2000 Inc., December 2000

2) Technologies Technical Framatome Technologies

2) Framatome TechnicalDocument "Stress Report 33-1201205-02, "Stress Document 33-1201205-02, Sununary for Report Summary for Reactor Vessel, Toledo Edison Company, Davis-Besse Unit No. I," SI File Reactor Vessel, Toledo Edison Company, Davis-Besse Unit No. I," 51 File W-ENTP-IIQ~ W-ENTP- 1Q-219P 219P

3) B&W Drawing 154628E, B&W Drawing Rev. 5, 1S4628E,Rev. S, "Closure Head Sub-Assembly,"

"ClosureHead Davis-Besse Document Sub-Assembly,"Davis-Besse Document Number M-503-212, SI File No. W-ENTP-I NumberM~S03-212,SI File No. W-ENTP-IIQ-219P IQ-219P Email from

4) Email

4) Prasoon Goyal from Prasoon Energy) to (First Energy)

Goyal (First Richard Bax to Richard Bax (SI), Thursday, September dated Thursday, (S1). dated September 20,2001, 4:4S AM, "Additional Design Input," Referencing B&W Drawing 154632E, 20, 2001, 4:45 AM, "Additional Design Input," Referencing B&W Drawing 154632E,Rev.

Rev.

2, 2, "Control Mechanism Housing,"

Rod Mechanism "Control Rod Document Number Davis-Besse Document Housing," Davis-Besse M-S03-213-2, SI Number M-503-213-2, S1 File No.

File No. W-ENTP- IIQ-219P W-ENTP-I1Q-219P

5) Framatome Document 51-5013435-02, Framatome Document SI-SOl34jS-02,"CRDMNozzJelBore Anal}sis," SI Dimensional Analysis,"

"CRDM Nozzle/Bore Dimensional SI File File No. W-ENTP-1 No. IQ-219P W-Eh"TP-IIQ-219P

6) Structural Integrity Calculation W-ENTP-07Q-301, Rev. 0, "CRDM Penetration J-Weld

6) Structural Integrity Calculation W-ENTP-07Q-301, Rev. 0, OlCRDM Penetration J-Weld Size Size Calculation" Calculation"

7) ASME

7) ASMEBoiler Vessel Code, Pressure Vessel and Pressure Boiler and 1989 Edition, Code, 1989 Section IIr, Edition, Section Appendices rn,Appendices 8)

8) Letter DBE-01-000133, Dated Le~terDBE-OI-000133, September 13, Dated September 2001 from 13,2001 Goyal (First Prasoon Goyal from Prasoon Energy) to (First Energy) to Dick Mattson Dick (S1), SI Mattson (SI), S1 File FileW -ENTP-II1Q-219P W-ENTP-1 Q-219P Email from

9) Email

9) Prasoon Goyal from Prasoon Energy) to (First Energy)

Goyal (First Richard Bax to Richard Bax (SI), Wensday, September dated Wensday, (S1), dated September 19, 2001, 12:39 PM, "Additional Design Input," Referencing B&W 19,2001,12:39 PM, "Additional Design Input." Referencing B&W Drawing 54628E Drawing 54628E "Closure Head "Closure Sub-Assembly," SI Head Sub-Assembly," 51 File No. W-ENTP-1 File No. IQ-219P W-ENTP-IIQ-219P Revision Revision 0o Preparer/Date PreparerlDate RLB 10-08-01 RLB 10-08-01 Checker/Date CheckerlDate STC 10-08-01 STC 10-08-01 File W-ENTP-l11Q-306 No. W-ENTP-1 Fi1e No. Q-306 Pae:e 13 Page 13 ofof 2424 Framalome ANP Framatome ANP Proprietary Proprietary NRC004-1241 NRC004-1241

Figure 1I - Finite Element Element Model Model Revision Revision o0

~

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rF~i":'""le~N~o-.

File No. -:::W~-:::ENT~P::--~l":'""l W-ENTP-l 1Q-306 ~3"':"06-:----'-----..=...t-------i..:P:-a-e-1-4-of-:--2-4---1 Page 14 of 24 Framatome ANP Proprietar,y Framatome Proprietary NRC004-1242 NRC004-1242

R= 100* [3]

Figure 2 - Top Hemispherical Head I Closure Flange Dimensions Revision o PreparerlDate RLB 10-08-01

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NRC004-1243 NRC004-1243

CRDM ClosurcH~d Bolt Holes (69 total [3]) (60 total [3])

6

,....'~----.- .. '-

17.180" 25.770" 34.360*

42.950" 5U40" Figure 3 - CRDM Penetration Locations

.~~

Revision o PreparerlDate RLB 1()"()8-O1 CheckerlDate STC 10-08-01 File No. W-ENTP-ll -306 P e 16 of 24 Framatome Framatome AN)P ANP Proprietary Proprietary NRCOO4.-1244 NRC004-1244

4.00" Attachment Eement Auachment Finite Element Zone to Modeled CRDM Tube Figure4a Figure 4a Figure Figure4b4b Reference 3 From Reference As-Modeled As-Modeled Revision o PreparerlDate RLB 10-08-01 CheckerlDate .STC lO-OS-Ol FileNo. W-ENTP-ll -306 Pa e 17 of 24 Framatome Framatome ANP Proprietary Proprietary NRC004-1245 NRC004-1245

Figure 5 - Applied Couples Attaching CRDM Tube to Hemispherical Head Revision o PreparerlDate RLB 10-08-01 CheckerlDate STC 10-08-01 File No.W-ENTP-ll -306 P e 18 of 24 Framatome FramatomeANPANPProprietary Proprietary NRC004-1246 NRC004-1246

ELEMENTS OCT 1 2001 TYPE NUM 13:44:22 Figure 6 - Applied SymmctIy Boundary Conditions Revision o PreparerlDate RLB 10-08'()1 CheckerlDate STC lO'()8'()1 File No. W-ENTP-ll -306 Pa e 19 of 24 Framatome FramatomeANP ANPProprietary Proprietary 1 0

NRC004-1247 NRC004-1247

Figure 7- Couples and Vertical 7 - Applied CoupJes Vertical Restraint at at Contact Surface (Lower Flange Contact Simulated with Gap Elements)

Revision Revision o _

{!. PreparerlDate Prarerate

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.~ ..' .

£LEMENTS TYPE NOH OCT 12001 14:11;$0 PRSS' PRES PRSS

£

• 1'05 Figure 8 - Applied NonnalOperating Pressure Load

.:::rPreparer,IDate RLB 10-0S-O 1

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ELEMENTS AN ocr 1 2001 TYPE JnIM 14:08:17

.PR£S 36378 Figure 9 - Applied Pressures To Simulate Closure Bolt Load Revision 0 Preparer/Date PreparerlDate RLB RLB 10-08-01

,.kerA/Date STC lO-Og-o1 Ch.~kerlDate STCIO-08-01

.l , .. '

File.No.

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ELDSENTS I\N OCT 1 2001 U:1t:09 F

t Figure 10 - Applied Gasket and Spring Loads Revision 0 PreparerlDate RLB lO.()8-O1 CheckerlDate STC lO'()8-O1 FileNo. W-ENTP-l1 306 Pa e 23 of 24 Framatome ANP Framatome ANP Proprietary Proprietary NRC004-1251 NRC004-1251

0 0 0 0 0

Figure - Locations Figure 11 .... Locations Where Gaps Are Closed Anywhere VertitaIPath Anywhere Along the Vertical Path of the Interference Interference Zone (Blue Lines)

Zone (Blue Lines) for for Worst Case Interference Values (Leakage Case Interference (Leakage Evaluation)

Revision o PreparerlDate RLB 10-08-01

• ~C~eckerlDate STC lO-Oj-Ol ".

q. ,.

FileNo. W-ENTP-l1 306 Pae 24 of 24 F,.amatomeAM' Framatome ANPProprietary Proprietary NRC004-1252 NRC004-1252

APPENDIX APPENDIX AA CRPDM CRDM to to Hemispherical Hemjmheri~aI Head Head Weld Weld Heights Resultine From ANSYS Resulting From ANSYS Innut FiJe WELD.INP Input File WELD.INP

{,

Revision 0

{r PreparerlDate CheckerlDate RLBIO..oS..ol STC 10..08..01 File No. W-ENTP-I1Q-306 Page Al of A3 Framatome FramatomeANPANPProprietary Proprietary NRC004-1253 NRC004-1253

~oo.

~

~020

.}Z' 028 8O1l4 31 032 tl 040 . 044

} 046

( }

9 43 Figure A-I Key Point Numbers Of Calculated Calculated Weld Heights Revision 0o Preparer/Date PreparerlDate RLB 10-08-01 Chkerat Ch~~erlDate STC......lO-O8'()1 STC 10-08.01 File No. W-ENTP-ll FileNo. W-ENTP-I lQ-306 -306 . Pa e A2 of A3 Page Framatome Proprietary FramatomeANP Proprietary NRC004-1254 NRC004-1254

The The following following table lists the table lists the height height ofofthe the interface interface between the hemispherical between the hemispherical head head inside inside surface surface (excluding the (excluding the clad) clad) and overall weld and overall weld diameter diameter for the tubes for the tubes (5.9712 (5.9712 inches from Section inches from 2.2.3). The Section 2.2.3). The height height values values are are for for the the point point onon the the weld weld circle circle nearest nearest the the top top dead dead center centerof ofthe the hemispherical hemispherical head bead and and aa height height at at the the farthest.

far1hest.

Also included Also included.in the table in the table is height at is height at the top of the top ofthe the weld, weld, which includes the which includes 3/1~il1ch the 3/16 inch weld weld butter.

butter.

The actual finite element model excludes the butter as part of the weld so the "Top of Weld Value" The actual finite element model excludes the butter as part of the weld so the "Top of Weld Value" minus the 33/

minus the /,616 inches inches was was used used in the final in the final finite element model finite element model to develop aa plane to develop that divided pJanethat divided the the modeled modeled CRDM CRDMtube tube structure structure along along the the top top of ofthe theweld.

weld. All All of ofthese these values valueswere were developed developed in in aa preliminary preliminary ANSYSANSYS input input file file named named WELDJNP WELD.lNP (includefmcJude on on the the project project CD-Rom).

CD-Rom).

Location From Height Height Measured Measured fromfrom Center Center of of Curvature Curvature of of Modeled Location From Modeled Hemispherical Head Hemispherical Bead .....

Tube Top-Dead Tube## Center Node l';"de Inside Inside Surface Surface Node Node Top of Weld TopoCWeld Center

1. 1. Height (in)

Hei2ht(lo) ## Heiglit Bei2bt(in) (in)

I1 Nearest Nearest 44 87.1989 87.1989 1004 1004 88.2614 88.2614 Nearest Nearest 66 86.08764 86.08764 1006 1006 87.15014 87.15014 3 Furthest 77 84.88764 1007 85.95014 Furthest 84.88764 1007 85.95014 Nearest Nearest 99 81.02136 81.02136 1009 1009 82.08386 82.08386 66 Furthest Furthest 10 10 78.37205 78.37205 1010 1010 79.43455 79.43455 Nearest Nearest 12 12 72.4916 72.4916 1012 1012 73.5541 73.5541 11 11 Furthest 13 68.11403 1013 Furthest 13 68.11403 1013 69.17653 69.17653 Nearest Nearest 17 17 86.76757 86.76757 1017 1017 87.83007 87.83007 15 15 Furthest 15 85.92749 1015 86.98999 Furthest 15 85.92749 1015 86.98999 Nearest Nearest 21 21 684.60745 "84.60745 1021 1021 85.66995 85.66995 22 22 Furthest 19 82.87501 1019 83.93751 Furthest 19 82.87501 1019 83.93751 Nearest Nearest 25 25 83.83298 83.83298 1025 1025 84.89548 84.89548 27 27 Furthest 23 81.87531 1023 82.93781 Furthest 23 81.87531 1023 82.93781 Nearest Nearest 29 29 80.57964 80.57964 1029 1029 81.64214 81.64214 31 31 Furthest Furthest 27 27 77.83217 77.8321'7 1027 1027 78.89467 78.89467 Nearest Nearest 33 33 79.73252 79.73252 1033 1033 80.79502 80.79502 39 39 .... Furthest 31 76.80169 1031 Furthest 31 76.80169 1031 77.86419 77.86419 Nearest Nearest 37 37 77.11485 77.11485 1037 1037 78.17735 78.17735 47 47 Furthest Furthest 35 35 73.6452 73.6452 1035 1035 74.7077 74.7077 Nearest Nearest 41 41 74.38134 74.38134 1041 1041 75.44384 75.44384 51 51 Furthest 39 70.3724 1039 71.4349 Furthest 39 70.3724 1039 71.4349 Nearest Nearest 45 45 " 73.44343 73.44343 1045 1045 1,74.50593 74.50593 S8 58 Furthest 43 69.25149 1043 70.31399 Furthest 43 69.25149 1043 70.3.1399 Nearest Nearest 49 49 70.54451 70.54451 1049 1049 71.60701 71.60701 63 63 Furthest 47 65.78542 1047 66.84792 Furthest. 41 65.78542 1047 66.84792 Revision Revision 00 V Preparer/Date PreparerlDate Checker/Date Checker/Date File FileNo.

No. W-ENTP-W-ENTP-l1 RLB RLB 10-080 STC 10-08-01 STC 10-08-01 10-08-01 I1Q-306 Q-306 E Page Page A3 A3 of of A3 A3 Framatome Framatome ANP ANPProprietary Proprietary NRC004-1255 NRC004-1255

APPENDIX B INTERFERENCE DIMENSIONS IJ\TfERFERENCE DIMENSIONS FOR FOR Davis-Besse, Unitl Unit!

FROM Framatome Document 51-5013435-02 Framatome Document Dimensional Analysis'"

Nozzle/Bore Dimensional 51-5013435-02 "CRDM Nozzle/Bore Analysis"

~ .

.~........

Revision o PreparerlDate RLB lO-O8"()1 CheckerlDate STC 10-08-01 File No. W-ENTP-IIQ-306 Page BloC B2 Framatome Proprietary Framatome ANP Proprietary NRC004-1256 NRC004-1256

fRA.ANp PROPRIETARY - 11-1013435-02

. . Page7of14 Tabl. 3: 8111'1l1'na1)'ofTop ami Bottom Dlmenslona! Ffta for CRDM Noz:zIeG In Ibe RV Closure Head for D-8 P.anMratlon DImeMIO"al FJt DImeIWIonaI fit Penetration DJmenslonal FIt Dlm* .,.ronalFJt

• Top (In) 8oItom (In)

• Top (In) Bottom lin) 1 ~~.0001) ,o.OO!~ ~ ~~.oooe) ~.OOO2J 2 (0.0019) (D.OO2D) 37 (O.D009) (0.0010) a (O.OQ13) (0.0015) 38 (O.oooe) (0.0007) 4 (0.0015) (D.OO1&) 39 (0.0007) (0.0008)

& (0.0007) (D.oOD9) ~ . (0.0003) (O.OO1Q) 6 (0.0012) (0.0008) ~1 (0.0006) (O.D0D9)

7 (0.0001) (0.0009) 42 (0.0009) (O.OOO5) 8 (O.OOO6) (0.0008) 43 (0.0009) (0.0009) 9 (O.ooon (O.oooe) 44 (0.0012) (0.0002) 10 (0.0013) (0.D002) 45 (0.0014) (0.0011) 11 (0.0009)* (0.0008) ~ (O.OOO4) (0.0012) 12 (0.0006) (0.0003) 47 (D.0013) (0.0007) 13 (0.0009) 0.0001 ~8 (0.0009) (0.0013) 14 0.0004 O.OOOS 49 (0.0002) 0.0002 16 (0.0007). (0.0006) 50 (0.0021) (0.0010) 16 (0.0009) (0.0009) 51 (0.0012) (0.0018) 17 (0.001.) (0.0013) 52 (0.0006) (0.0009) 18 (0.0009) (Q.OOO7) 53 (0.0016) (0.OD1!)

19 (D.ooo7) (0.0004) ,.64 0.0000 0.0001 20 (0.0008) (0.0008) as (0.0011) (0.0010) 21 (0.0002) 0.0001 56 (0.0016) (0.0011) 22 (0.0004) (o.ooos) 57 (0.0010) (O.OOCI6) 23 (0.0014) * (o.ooD6) A * .(0.0008) (0.0005) 24 (0.0016) 0.0004 59 (0.0008) (0.0001)

'26 (0.0007) (0.0012) eo (0.0005) (0.0011) 2.6 (0.0006) (0.0008) 81 (0.0012) (0.0003)

'D (0.0011) (O.DDOS) 62 (0.0013) (O.OOO4) 28 (0.0004) (0.0008' 63' (0.001") (0.0015) 29 (0.0009) (0.0010) 84 (0.0007) (0.0005) 30 (0.00f3) f55 (0.0010)

. 31 (D.OOO8)

(O.oof1)

(0.001D) 6B (0.0011)

(0.0004)

(0.0012) 32 (0.0007) 0.0002 ff1 (0.0005) (0.0006) sa (0.0018) (0.0003) 88 (0.0012) (0.0013) 34 . (0.0016) (0.0010) 89 (0.0010) (0.0005) .

35 (0.0002) 0.0010 Revision 0 l) PreparerlDate RLB 1O-OS-O 1 CheckerlDate .. 'STC 10-08-01

~

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NRC004-1257 NRC004-1257

APPENDIX APPENDIX CC CASE CASE STUDY STIJDY FOR FORUSE USE OF OF CONTAC52 CONTACS2 ELEMENTS ELEMENTS I q TO TO IMPOSE IMPOSE T7 ii *

• INTERFERENCE INTERFERENCE , FIT LOADS FIT LOADS Revision o PreparerlDate RLB 10-08-01 Ch~k~lDse STCIO-08-01 Framatome FramatomeANP ANPProprietary Proprietary NRC004-1258 NRCOO4--1258

A comparison study was performed software package [1]

perfonned using the ANSYS software verify that the

[ I] to veritY CONTAC52 CONTAC52 clement element would be an adequate means of simulating the interference interference load between between the CRDM housing bousing tube and the closure c10sure head. The verification oftwo finite consisted oftwo verification consisted element models.

The first would use the CONTCONTAC52AC52 gap element while the other would use a simple simple application of of imposed displacement.

imposeddispJacement.

Model #1 - CONTAC52 CONTAC52 Elements The first model was a simple tube and plate model model as shown in Figure C-J. C-1. The plate was 20"x20"x4" and was assumed to be infinitely rigid (Modulus of Elasticity, E ==

20"x20"x4" 30e12 psi).

= 30e12 A 4 inch diameter hole was centered through the plate into which was inserted a 3.998 inch diameter, 10 inch long tubetllbe that had a waD of0.4999 inches. The slight reduction in tube wall thickness of interference function CONTAC52 elements. The tube was diamrter was necessary to support the interference diaJljeter that its base was flush with the~ttom inserted such tbatitsbase ving 6 inches of the tube the bottom of the plate kIc. *ving protruding from the top. The tube was protruding modeled with a Modulus cf wasmodele<f of 30e6 psi.

Elasticity. E, of30e6

(:f Elasticity, CONTAC52 elements were applied at the plate to tube interfacl. interference value of -0.01 interfacL with an interference -0.01 between the tube and interference between inches. This should simulate the existence ofa 0.01 inch radial interference the plate throughout the circumference circumference of the tube and the 4 inch thickness thickness ofthe plate.

The plate edges and bottom along with the bottom of the tube are held with synunetry symmetry boundary conditions. SeeANSYS See ANSYS input file TEST2.INP on the project CD,;,Rom. CD-Rom.

The resulting resulting stress intensity in the tube is shown in Figure C-2 and peaks at 233629 psi.

Model #2 #2 - Imposed Displacements Imposed Displacements The second model models only the tube from the previous model. Dimensions, mesh density and materials are all the same. The interference interference load forthls analysis consisted of a series for this analysis series of imposed displacements located at the same locations as the gap elements

-0.01 inch radial displacements elements in the previous analysis.

The base of the tube was held with symmetric symmetric boundary boundary conditions and a pair of opposing nodes circumferential direction to prevent rigid body motion. Figure C-3 showns the was held in the circumferential boundary conditions resulting model and the boundary (mcluding the imposed-0.01 conditions (including imposed -0.01 inch radial displacements). See ANSYS input file TEST2a.INP displacements). TEST2aINP on the project CD-Rom.

The resulting stress intensity for this load is shown in Figure C-4 and peaks at 233776 psi. The 0.06290.4 greater than the theoretical same load applied via stress intensity for this load method is 0.0629%

CONTACS2 elements.

the CONTAC52 interference loading was CONTACS2 element for interference elements. Clearly use of the CONTAC52 was acceptable.

acceptable.

Revision o PreparerlDatc RIB 10-08-01

,'." J ChecksrLDatc STC 10-08-01 FHe No. W-ENTP-l1Q-306 Pae:e C2 of C6 FramalomeANP Framatome ANP Proprietary Proprietary NRC004-1259 NRC004-1259

ELEMENTS AN.* I AUG 21 2001 MAT NUM 14.00.32 Figure C-l - Finite Element Model Using Gaps Revision 0 PreparerlDate RLB 10-08-01 CheckerlDate STC 10-08-01

• File No. W-ENTP-l1 -306 P ee3 of C6 Framatome Framatome ANP Proprietary Proprietary NRC004-1260 NRC004-1260

NOIlAL SOLt1l'ION AN AUG 21 2001 STEPel 13:38:4)

SUB .1 TIME-l SINT IAVG)

DMX *. 014065 SMN .1700 SMX .233629

,ga  ::maYO,;. CitHE ,:!t.~~"'!. p:a:::::.~--~

1700 53240 104779 156319 207859 27470 79009 130549 182089 233629 FigureC Stress Intensity in Tube for Gap Interference Analysis Revision o

~ PreparerlDate RLBl()'()8'()1

". " ". V . ,. .

C=:':h-e-=ck-er.-:m=-.-at-e+s.....'{1--C-I-O'O-S.()-l+-------+---------+-.........-.,....--~-I FileNo.W.;ENTP-U 306 Pa e C4 of C6 Framatome FramatomeANP ANP Proprietary Proprietary

---

.... -~---.- ... -

NRC004-1261 NRC004-1261

I\N AUG 21 2001 14:1':43 Figure C Finite Element Model with Boundary Conditions Using Imposed Displacements for Loading Revision o PreparerlDate RLB 10-08-01 CheckerlDate STC 10-08'()1 File No. W-ENTP-ll -306 Pa eCS of C6 Framatome Framatome ANP ANPProprietay Proprietary NRC004-1262 NRCOO4--1262

NODAL NODAL SOLUTION AN I\N AUG 21 2001 STEP-11 STEP. 13:=51:=31 13:5.1:31 SUB -1 al TIME-1 TIME-I SIm SINT (AVO)

(AVG)

DMX u.014072 DMXV-.014072 S1ON W1701 SMN -1701 S!CC .233776 SMX .233776 SMXB-26997S SMXB.269975

. .*6* !Gd! ;:S:. et4i#P e.*=~

1701 53273 104845 104945 156418 207990 27487 2?487 79059 79059 130632 182204 182204 233776 Figure C Stress Intensity in Tube for Imposed Displacement Analysis 0 Revision "Revision o 0

SPreparer/Date PreparerlDate Checker/Date CheckerlDate RLB 1()'()8"()1 RLB 10-08-01 src 1o-os0-1 STC 10-08"())

File No. W-ENTP-1 W-ENTP-l1IQ-306 306 Page Pa C6 of C6 Framatome Framatome ANP Proprietary Proprietary NRC004-1263 NRC004-1263

APPENDIX APPENDIX D ADDITIONAL EVALUATIONS ADDITIONAL EVALVAnONS OF CRDM CRDM #3 GROUP GROup INTERFERNCE CONDMONS INTERFERNCE CONDmONS o

VParer/Date Revision 0 PreparerlDate RLB lO-OS'()l Checker/Date RLB 10-08-01

• CheckerlDateSTCSTC 10-08-01 FileNo. W';ENTP-lllQ-3061 File No. W-ENTPl -306 P e DJ Page DI of D3 FramatomeANP Proprietary Framatome Proprietary NRC004-1264 NRC004-1264

~~~--~--------~--~------~--------------------------~-------,

Objective D.l.0 Objective D.1.0 the original In the evaluation, CRDM origina1 evaluation, CRDM Tube Tube 33 was fucluded in the was included the finite eltmlent model. Per Table finite element Table I1 of of 5.6. CRDM Section 5.6, Section CRDM Tube was grouped Tube 3 was CRDM Tubes grouped with CRDM 2, 4 and Tubes 2,4 and 5. Also in S. Also in Table l,CRDM Table 1,CRDM Tube 3 was Tube.3 actually loaded was actually with CRDM loaded with CRDM Tube 2's interference values, Tube 2's interference Tube 2 having been selected baving been selected astbe as conservative interference the most conservative interference condition condition of oftbe group for leak the group rate evaluations.

leak rate evaluations.

It was shown in TableTable 3, Section Section 6.1, 6.1. that that the Tube 33 had no gapsgaps at the very bottom of the interference interference zone, resulting in no zone,resuJting no potential potentia1 leakage.

leakage. As a result, result. additional additional evaluations eva1uations were were performed in performed in order investigate investigate the other tubes tubes in the group.

D.2.0 Finite Fmlte Element Element Model Model Changes Changes The only changes that will the original will be made to the original finite model is to Tube 3's fmite element model interference 3 'sinterference D-lIi~s values. Table D-I lists the affected tube group group interference interference values.

values. Interference Interference values values are from from ReferenceS Reference5 (see also Appendix B). B).

TableD.;1 Table D-1 Interference Values For CRDM Ioterferf:oce CRDM Tube Tube 33 Group Group Tube Diametrical Interference Dimensions Diametrical Interference Dimensions (in) (In)

Tube Tube___.Top Top Bottom 2 0.0019 0.0019 0.0020 3 0.0013 0;0013 0.0015 0.0015 4 1 0.0015 0.0015 0.0015 5 0.0007 ' < 0 . 0 00.0009 09 evaluation used Tube 2 interference The original evaluation interference va1ues. additional evaluations will be values. Two additional be performed. The fITst first will use Tube 4 interference interference values andis and is conservatively considered considered to include include Tube 3 since the area of concern is the bottom portion of the interferenceinterference zone. The second evaluation will wi1J use Tube 5 interference values. The modified ANSYS ANSYS input files for these two evaJua:tionsare evaluations are named DB4CRDM.INP DB4CRDM.1NP and DB5CRDM.INP DB5CRDMJNP (both are included included on the project CD-ROM).

D.3.0 Gap Opening Opening Evaluation Evaluation Results For the evaluation interference values.

evaluation using Tube 4 interference values, it was once again determined that there were no no gaps at the very bottom ofthe interference interference zone (similar (similar to the original evaluation results). The gap data for the evaluation is included in the Excel spreadsheet spreadsheet APPENJ).xLS D.XLS (included on the project CD-Rom).

Revision o

~

~

PreparerlDate .RLB 10.:08-01

....CheckerlDate STC 10-Og.,o1 File No. W-ENTP-t 10-306 Page D20f D3 FramatomeANp Proprietary FramatomeANP Proprietary

'.-~~".-'

NRC004-1265 NRC004-1265

1*

For the evaluation using Tube 5 interference detennined that several gaps were now interference values, it was determiined through the interference available through interference zone.

pp opening along the vertical Table D-2 lists the smallest gap modeled Tube 3 for both vertical path of modeled interference cases described in the previous interference previous page. In the case of the Tube 4 interference interference case where interference value is listed in 1he there was no gap, the minimum interference of a negative the form ora negative value. A complete list of tube results for the Tube 4 and 5S cases can be found complete found in the Excel spreadsheet spreadsheet APPEND.XLS APPEN_DJUS (included on the project project CD-Rom).

Table D-2 TableD-2 Minimum Gap Results for Tube 3 Leakage Evaluations Evaluations Interference Interference Value Used VatueUsed Minimum Gap (iDches)

MInimum (inches)

Tube Number Number .......... _,

2 (Original (Ozigina1 Case) .4.30002483 (Interference)

-r:;)0002483 4 -'j.OOO00942

-"j.00000942 .(Interference)

(lnterference) 5 0.000074799 0.000074799 D.4.0 Conclusions D.4.O Conclusions Based on the results of these additional evaluations, Based evaluations, it bas has been determined that Tubes 1 1,2,

,2, 3 and 4 provide no gap through whicbleakage which leakage may occur during normal operating conditions.

operating conditions. All other tubes have gaps through which leakage may occur.

Revision 0o Preparer/Date PreparerlDate o0-o0-01 RLB lO-:OS"()l RLB Checker/Date STC 10-08-01 File No. W-ENTP- IQ-306 Page D3 of D3 Framatome FranatomeANP Proprietary Proprietary NRC004-1266 NRC004-1266

NRC LETTERS -REVIEW NRC - REVIEW AND APPROVAL APPROVAL REPORT

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The NRC Letters* Approval Report (ED 7159-7) should be co~pleted Letters - Review and Approval completed by.the by the Regulatory Affairs Affillrs Section, Section.

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correspondence. This summary

.should should identify ifif the correspondence correspondence is in response to any previous correspondence and why previous correspondence why the letter i being written.

BLOCK 4 COMMITMENT LIST ADDED TO LETTER - Preparers COMMITMENT Preparers checks checks the block to indiate a commitment list hal; a commitment has been included included with the letter.

BLOCK 5 PERIODICINON-PERIODIC PERIODIC/NON-PERIODIC REPORT REPORT - Identify whether whether this correspondence correspondence is a Periodic Periodic or Non-Periodic Report as identified identified in Nuclear Group Procedure Procedure NG-NS-00807.

NG-NS-00807.

BLOCK 6 DATE RESPONSE RESPONSE DUE TO BE SUBMITTED SUBMITTED TO NRC -- Initiator enters the date the correspondence correspondence is due to the NRC~

NRC. IfIf the correspondence correspondence does not have a required due date, the block.shall block shall be marked not applicable (NA).

applicable (NA).

BLOCK 7 SPECIAL HANDLING-Initiator HANDLING - Initiator checks if the correspondence distribution to the NRC. 1.

correspondence requires special distribution I~

yes, the Regulatory Affairs clerk enters date date the correspondence sent.

correspondence is .sent.

BLOCK 8 PREPARED BY - Initiator enters the names of individuals responsible for providing technical information BLOCK 8 PREPARED BY - Initiator enters the names of individuals responsible for providing technical information for the correspondence correspondence along with his/her name.

BLOCK 9 NOTARY -- Initiator checks ifif the correspondence correspondence is required to be notarized.

to BLOCK 10 LICENSE FEE REQUIRED - Initiator checks if a license fee is required, per the requirements requirements of Part 170.

170. IfIf yes, the initiator shall complete a Voucher Voucher Check Check Authorization (Form 294)' 294) and obto appropriate appropriate fees to accompany accompany the correspondence. correspondence.

BLOCK 11 BLOCK 11 ADDITIONAL REFERENCES ADDITIONAL REFERENCES - Initiator Initiator entersenters any additional NRC correspondence correspondence or documents documents that pertain to the subject subject'correspondence.

correspondence.

BLOCK 12 12 COMMITMENT COMMITMENT NO(S). CLOSED - Initiator enters the COmmitment Commitment Management System number(s)

Management System number(s) of any commitments that are closed by the subject subject correspondence. .

BLOCK 13 BLock COMMENTS - Initiator COMMENTS InHiator or any reviewer enters appropriate comments regarding the subject correspondernce.

corresponderice~

BLOCK 14 14 REVIEW AND APPROVAL APPROVAL - Initiator checks and and lor/or enters the desired desired reviewer(s).

reviewer(s). The technical accuracy accuracy of a response response to *the -the NRC is the responsibility respbnsibility of the Director Management individual Director and Management assigned assigned the action. .

BLOCK 15 DATE ADDED BY - Distributor checks the Date Added to Letter Block and signs the Date Added By By block to indicate indicate the original letter has been dated prior to distribution distribution to the NRC. .

BLOCK 16 . DIstribution to the NRC shall be made DISTRIBUTED BY - Distribution made by the Regulatory Affairs Section. Distributo signs the Qistributed Distributed By block and completes completes the Date Sent to NRC block.

BLOCK 17 ADDITIONAL DISTRIBUTION - Initiator enters ADDITIONAL DISTRIBUTioN enters individuals distribution that are not individuals requiring distribution not on the the standard distribution distribution list. ".

BLOCK 18 DISTRIBUTED BY - Distributor signs the Distributed by block and completes the Date of Blind..

BLOCK 18 DISTRIBUTED BY.- Distributor sigos the Distri~uted 'By block and completes the Date of Bnnd~, .

Distribution block. . .

' ......... J " . "" ....... # w' o SII-0059 9 S11-00599 NRC027-1699 NRC027 -1699

~~~

R sC. c.15~151 DOCKETED DOCKETED USNRC USNRC ArstEnergy FArstEnergy September September 9, 9, 2009 2009 (11 (11 :OOam)

OOam)

Davis-Besse Davis-Besse Nuclear Power Station Nuclear Power 5501 North State Oak Harbor, Station State Route 2 Ohio 43449-9760 Harbor, Ohio 43449-9760 OFFICE OF OFFICE OF SECRETARY SECRETARY

~l G. Campbell campbell RULEMAKINGS RULEMAKINGS AND ADJUDICATIONS ADJUDICATIONS STAFF AND STAFF 419-321-8588 419-321-8588

~ President- Nuclear President Nuclear Fax: 419-321-8337 Fax: 419-321-8337 Attachment Contains Contains Restricted Material Per 10 CFR 2;790 Material 2.790 Docket Number 50-346 U.S.

U.S. NRC In NRC re DAVID L j4 I' f' GEISEN ,7lc\.. ti EXIbltW F'1 #......

EJChIbIt, 3

,,{,;;::;..-_

Inre DAVID GEISEN License Number NPF-3 Docket # 1 A-05-052 1A-05-052 Date Marked-for Marked-for ID:l1J.L.

ID 2008 (Tr. p.&Z '5 (Tr.p.*' )

Serial Number Serial Number 2744 October 30130, 2001 Date Offered inEv:

Offered in J1.Ji-.

.L.. 2008 (Tr. p 82- '" )

2008 (Tr. p.

Witnesslpanel:_.l..rJ-J,l..:.iX-----

Through Witness/Panel: .

Acio: ~ REJECTED Action: REJECTED WITHDRAWN WITHDRAWN U.S. Nuclear Nuclear Regulatory Commission Attention: Document Document Control Desk Date: .Ll4L-.

De:14. .2008 2008 (Tr. P. §ilz ,1..0 ))

(Tr. P~.L Washington, D.C. 20555-0001

Subject:

Transmittal Transmittal of Results of Reactor Pressure Vessel Head Control Rod Drive Drive Mechanism Nozzle Penetration Visual Examinations Mechanism Examinations for the Davis-Besse Nuclear Power Station Ladies and Gentlemen:

During a public meeting between the Davis-Besse Davis-Besse NuclearNuclear Power Station (DBNPS) staff and the Nuclear Nuclear Regulatory Commission Commission (NRC) staff on October October 24, 2001, 200 1, concerning concerning the FirstEnergy Nuclear Operating Company (FENOC) response (letter Serial Number Nuclear Operating 2731, dated September 4,2001) 2731, 4, 2001) to NRC Bulletin 2001-01, 2001-01, "Circumferential "Circumferential Cracking Cracking of Reactor Pressure Vessel Head Penetration Penetration Nozzles,"

Nozzles," the DBNPS staff committed committed to provide provide pictorial documentation of the visual examinations of the reactor pictorial documentation reactor pressure vessel head performed during the DBNPS 10 th , 11th 1Oth, 11h and 12th refueling refueling outages. This documentation documentation is provided in Attachment 1 as the DBNPS report, "Results "Results of Visual Examination Examination of Reactor Reactor Head CRDM CRDM Nozzle Penetration Performed Performed During During 1996, 1998, 1998, and 2000."

This report is considered to be restricted by the FENOC and is requested requested to be withheld from public disclosure pursuant disclosure pursuant to 10 CFR 2.790. An affidavit complying with the requirements requirements of 10 CFR 2.790 is provided in Attachment Attachment 2 citing the basis for this report to be withheld from public disclosure.

The inspections inspections performed performed during the 10 th , 11th, 1O0h, 11 th, and 12 th Refueling 12th (10RFO, Refueling Outage (lORFO, conducted conducted April 8 to June 2, 1996; 11RFO, conducted lIRFO, conducted April 10, to May 23, 23, 1998; and,

\. ýTrý ý

,1ý Ný 1__fýeT e_; s.cZý -blv& -Tý 4; Cýk

Docket Docket Number Number 50-34650-346 License Number License Number NPF-3 NPf-3 Serial Serial Number Number 2744 2744 Page Page 2 of of2 2 12RFO, 12RFO, conducted conducted AprilApril 1 to to May M~y 18, 2000) consisted 18,2000) consisted of of aa whole whole headhead visual visual inspection inspection of of the the RPV head head inin accordance accordance with with the DBNPS DBNPS Boric Boric Acid Acid Corrosion Corrosion Control Control Program Program pursuant pursuant to to Generic Generic Letter Letter 88-05,

~8-05, "Boric "Boric Acid Acid Corrosion Corrosion of Carbon Carbon SteelSteel Reactor Reactor Pressure Pressure Boundary Boundary Components Components in in PWR PWR Plants."

Plants." The The visual visual inspections inspections were were conducted conducted by remote camera and included below insulation by remote camera and included below insulation inspections of inspections of the the RPV RPV bare bare head head such

-that that thethe Control Control Rod Rod Drive Drive Mechanism Mechanism (CRDM) (CRDM) nozzlenozzle penetrations penetrations werewere viewed.

viewed.

During During 1ORFO, lORFO, 65 65 of 6969 nozzles nozzles werewere viewed, viewed, during during 111RFO, IRFO, 50 50 of of 69 69 nozzles nozzles were were viewed; viewed, and during during 12RFO, 12RFO; 45 45 ofof 69 69 nozzles nozzles werewere viewed.

viewed. It It should should be noted noted thatthat 19 19 of of the the obscured obscured nozzles nozzles in in 12RFO 12RFO were were also also those those obscured obscured in in 11 RFO. Following l1RFO. Following 11 RFO, llRFO, the the RPV RPV headhead was was mechanically mechanically cleaned cleaned in in localized localized areas areas as as limited limited by by the the service service structure structure design. Following12RFO, Following .12RFO, the the RPV RPV headhead was was cleaned cleaned with demineralized demineralized water water

...... ,. ..... " to the extent possible to provide the extent possible to provide:a clean head a clean head for evaluating future inspection evaluating future inspection results . results.

The The affected affected areas areas of of accumulated accumulatec(boric boric acid acid crystal crystal deposits deposits were were video video taped, taped, and and have have subsequently subsequently been been reviewed reviewed -with with specific specific focus on on boric boric acid acid crystal crystal deposits deposits withwith reference reference to to the the CRDM CRDM nozzlenozzle penetration penetration leakage leakage as as previously previously observed observed at at the the Oconee Oconee Nuclear Nuclear Station, Station, Unit Unit 3 (ONS-3)

(ONS-3) and and at at Arkansas Arkansas Nuclear Nuclear One, One, Unit Unit 11 (ANO-1).

(ANO-I). DuringDuring the 12RFO inspection, 24 of the 69 nozzles were obscured by boric the 12RFO inspection, 24 of the 69 nozzles were obscured by boric acid crystal deposits acid crystal deposits that that were were clearly clearly attributable attributable to to leaking leaking motor motor tube tube flanges flanges from from thethe center center CRDMs.

CRDMs. A A further further subsequent subsequent review review of of the the video video tapes tapes has has been been conducted conducted and and the the results results of of this this review review did did not not identify identify anyany boric boric acid acid crystal crystal deposits deposits that that would would havehave been been attributed attributed to to leakage leakage fromfrom thethe CRDM CRDM nozzlenozzle penetrations, penetrations, but but were were indicative indicative of of CRDM CRDM flangeflange leakage.

leakage.

The The aforementioned aforementioned video video taped taped images images of of areas areas ofof accumulated accumulated boric boric acid acid crystal crystal deposits deposits havehave been been converted converted to to photographic photographic imagesimages andand are are contained contained in in the the attached attached report.

report.

If If you you have have anyany questions questions or or require require further further information, information, please please contact contact Mr.

Mr. David David H. H.

Lockwood, Manager-Regulatory Affairs Lockwood, Manager-Regulatory Affairs at (419) 321-8450. at (419) 321-8450.

Very truly Very truly yours, yours,

~vVV\

Enclosure Enclosure Attachments Attachments cc:

cc: J.J. E.

E. Dyer, Dyer, Regional Regional Administrator, Administrator, NRC NRC Region Region III ill S.S. P. Sands, DB-1 P. Sands, NRCINRR Project DB-1 NRC/NRR ProjectManager Manager D.

D. S.S. Simpkins, Simpkins, DB-1 DB-l Acting Acting Senior Senior Resident Resident Inspector Inspector Utility Radiological Safety Utility Radiological Safety Board Board

Docket Number 50-346 Docket 50-346 License Number NPF-3 License Number Serial Number Serial Number 2744 Enclosure Page 1 of 1 SUPPLEMENTAL INFORMATION SUPPLEMENTAL INFORMATION RESPONSE TO IN RESPONSE TO NRC BULLETIN BULLETIN 2001-01 FOR FOR DAVIS-BESSE NUCLEAR DAVIS-BESSE NUCLEAR POWERPOWER STATION UNIT NUMBER UNIT NUMBER 1 This letter is submitted pursuant to 10 CFR 50.54(f) and contains CPR 50.54(0 contains supplemental information concerning information response (Serial 2731, concerning the response September 4, 2001) 2731, dated September 200 1) to NRC Bulletin 2001-01, "Circumferential Cracking 2001-01, "Circumferential Cracking of Reactor Pressure Vessel Head Reactor Pressure Penetration Nozzles," for the Davis-Besse Penetration Nozzles," Nuclear Power Station, Unit Number 1.

Davis-Besse Nuclear I, Guy G. Campbell, state that (1) I am Vice President - Nuclear of the FirstEnergy Nuclear Nuclear Operating Company, (2) I am duly authorized to execute and file this Operating Company, certification on behalf of the Toledo Edison Company and The Cleveland Electric Illuminating Company, and (3) llluminating statements set forth herein are true and correct to the (3) the statements best of my knowledge, information and belief.

B'-~

Guy GGI Campbell, Vice President Nuclear subscribed before me 30th day of October, Affirmed and subscribed 2001.

October, 2001.

State of Ohio - Nora L. Flood Notary Public, State My commission expires September September 4, 2002.

4, 2002.

Docket Number 50-346 License License Number NPF-3 Number NPF-3 Serial Number 2744 Attachment 11 Attachment Page 1 of I1 Davis-Besse Nuclear Power Station Results of VjsuaI Examjnation of Reactor Head CRDM Nozzle Penetrations Visual Examination Penetrations in 1996, 1998, and 2000 Performed in*1996, Perfonned (46 Pages Follow)

.1 j)

/

Davis-Besse NPS Davis-Besse A

Results of Visual Examination Examination J

of of Reactor Head CRDM nozzle nozzle penetrations penetrations Performed in 1996, 1998, and 2000

,* Arrangement - 69.CRDM Nozzles Nozzle Arrangement Nozzles

.1{

SllJD HOLE I<<JWIl~RS 11 THRI.I 60 I sa Ct .

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84*1'/.~J)IA. BOLT HOLtS 0+1 1 'I4Yr" IHA. B.C.

I~OT PRESSURE RETAINING iI BOLtlNGI Y -. RATRVESSEL

.....R~ACT~ VESSEL .CLOSUREIHEAD CLOSURE _

- ....... _.. HEAD KEY PLA" KET PLAN FENOC RESTRICTED RESTRICTED INFORMATION INFORMATION

A ". ",,> .,," ~, ,~ e . .'A' " , ", ".' . " ~ ,

,~'., .~

-~ ~

j Nozzle Core Quadrant 11996 1996 Inspection Inspection results results 2000 Nozzle Core Quadrant 1998 1998 Inspection Inspection results results 2000 Inspection Inspection results results

,*No. Locat.

j No. Locat.

,~

See Note Note 1.0 1.0

,~

I 11 H8 H8 1 FlangeLeak Evident Flange Evident Flange FlangeLeak Evident Leak Evident I IH 22 G7. FlangeLeak Evident Evident FlangeLeak Evident Flange Evident

~~

4 Flange 33 G9 i

it' 44 ,Kg

-K9 K7 ~

1 2

FlangeLeak Flange Flange Leak Evident FlangeLeak Leak Evident Flange Leak Evident Leak Evident Evident FlangeLeak Flange FlangeLeak FIC1nge Leak Evident Evident Leak Evident Evident Leak Evident Flange Leak Evident I

5 3 Flange 66 F8 Fa 1 Flange Leak Evident Flange Flange LeakLeak Evident Evident IG 77 H1O H10 2 Flange Leak Flange Leak Evident Flange Leak

'Flange Leak Evident I

l a8 L8 La 3 Leak Observed No Leak 'No No Leak Observed q

g9 H6 H6 4 Leak Observed No Leak Observed iNo Leak

'No Leak Observed Observed

"t 10 10 F6 F6 *4 Leak Observed No Leak Observed Leak Observed No Leak Observed 11 11 Fl10 F10 1 Flange Leak Evident ,Flange Flange LeakLeak Evident Evident

~~ 12 Li0 Leak Observed No Leak Leak Observed No Leak Observed 2

i 12 L10

~ 13 13 14 14 L6 L6 E7 E7 I 3 4

No Leak Recorded Leak Recorded Flange Leak Evident No Leak Observed Evident Flange Leak E;vident I I 15 15  : E9 E9 1 Flange Leak Evident Evident Flange Leak Evident

.. 16 16 . Gll G11 1 Flange Leak Evident Flange Leak Evident i 17 17 .. Ki11 K11 2 No Leak Observed 'INo No Leak Observed

• 18 18 M9 M9 2 Recorded No Leak Recorded No Leak Observed I 19 19  ! M7 M7 3 No Leak Observed Recorded No Leak Recorded 20 20 KS K5 3 No Leak Observed No Leak Observed 21 21 G5 GS 4 No Leak Observed No Leak Observed 22 DS , Flange Leak Evident Flange Leak Evident 22 08 1 23 23 .. H12 H12 2 No Leak Observed Observed No Leak Observed Observed 24 24 NB N8 3 No Leak Recorded Recorded No Leak Recorded Recorded I .j 25 25 H4 H4  ! 4 No Leak Recorded Recorded Observed No Leak Observed 26 26 ~

ES .

ES 4 No Leak Recorded Recorded No Leak Observed

'No Observed K:l 27 27 Eli i E11 1 .Flange Flange Leak Evident Evident Flange Leak Evident Evident 28 28 -, mil M11 \ 2 No Leak Recorded No Leak Observed Observed 29 29 MS M5 3 No Leak Recorded No Leak Observed Observed 30 30 D6 4 No Leak Observed Observed No Leak Observed Observed

• 31 Dl10 010 " 1 Flange Flange Leak Leak Evident Evident Flange Flange Leak Evident Leak Evident 32 32 Fl12 F12 1 Flange Flange Leak Leak Evident Evident Flange Flange Leak Leak Evident Evident 33 33 Li2 L12 2 No Leak Recorded Recorded No Leak Observed No Leak Observed 34 34 N10 N10 2 'No No Leak Recorded Recorded No Leak Observed Observed 35 NS N6 3 No Leak Leak Recorded Recorded No Leak Leak Recorded Recorded 36 36

! L4 L4 3 No No Leak Recorded Recorded No No Leak Observed t,

37 38 38

~ F4 F4 C7 C7 4

4 No Leak Leak Recorded No Leak Recorded Leak Recorded Recorded No Leak Observed

,Flange Flange Leak Observed Leak Evident Evident

, 39 09 C9 1 Flange Flange Leak Leak Evident Evident Flange Flange Leak Leak Evident Evident

. ' 40 40 G13 G13 1 Flange Flange Leak Leak Evident Flange Flange Leak Leak Evident Evident 41 41

~

K1 K13 3 . 2 No No Leak Leak Recorded Recorded No Leak

'No Leak Observed Observed 42 42 09 09 2 No No Leak Leak Recorded Recorded No No Leak Leak Recorded Recorded 43 43 07 07 3

• No No Leak Leak Recorded Recorded No No Leak Leak Recorded Recorded

~ 44 44 K3 K3 3 No No Leak Leak Recorded Recorded !No No Leak Leak Observed Observed 45 45 G3 G3 4 No Leak Leak Recorded Recorded No No Leak Leak Observed Observed

,i\1

• 46 46 D4 04 4 No No Leak Recorded Leak Flange Recorded No No Leak Leak Observed Observed A 47 47 D12 D12 1 Flange Leak Leak Evident Evident Flange Flange Leak Leak Evident Evident" FENOC RESTRICTED FENOC RESTRICTED INFORMATION INFORMATION

• ý-:* W N - ..... ,

A 1111 , .:" *  : -:* * ,* . , ., , . .. . ~":. ... ' , ' " , ,,' I': ....

- "'" 1. ,',,- ,,~ ......

711 .' .~

11111 .. " " ~

Nozzle Nozzle Core Core Quadrant 1996 Inspection results results 1998 1998 Inspection Inspection results 120002000 Inspection Inspection results results No. Locat.

. 4~ ~

I A*4Z.

48 48 N12 N12 2 No No Leak Recorded Recorded No Leak Observed No Observed 49 N4 N4 3 No Leak Recorded No Recorded No Leak Observed No 50 C5 C5 4 No No Leak Recorded Recorded No Leak Observed 51 cil C11 1 Flange Leak Evident Flange Evident Flange Leak Evident I

~

52 53 E13 E13 M13 M13 1

2 No Leak Recorded No Recorded No Leak Recorded Recorded Flange Leak Evident No Leak Leak Observed Observed Evident I?l

!6 54 Oil 011 2 No Leak Recorded No Leak Leak Observed III 55 56 57 05 ~

M3 M3 E3 E3 B8 ~,

I 3

3 4

No Leak Recorded No Leak Recorded No Leak Recorded Leak Recorded Recorded

. No Leak Recorded No Leak No Leak Recorded Leak Observed Leak Observed I

58 88 1 No Leak Recorded Recorded .; Flange Leak Evident Evident It 59 H14 H14 2 No Leak Recorded No Leak Observed Observed

~l

~i 60 PS P8 3 No Leak Recorded . No Leak Recorded Recorded i 61 61 H2 H2 4 No Leak Recorded No Leak Observed I

62 B6 4 No Leak Recorded No Leak Observed

~ 63 BID 810 1 No Leak Recorded Recorded Flange Leak Evident

~ 64 Fl14 F14 1 No Leak Recorded Recorded Flange Leak Evident g

I~t

~1 65 65 66 67 Li4 l14 Plo Pi0 P6 P6 ~

2 2

3 No Leak Recorded Recorded No Leak Recorded Recorded No Leak Recorded

! No Leak Observed

,No No Leak Recorded No Leak Recorded

~ 68 L2 3 No Leak Recorded No Leak Observed Observed

~

~.

~4.

69 F2 F2 4 No Leak Leak Recorded Recorded No Leak Leak Observed Observed

~ -. ~

Filed as h/RCS hlRCS leakageleakage issues/nozzle issues/nozzle review Table Table Notes:

1 In 1996 during 1010 RFO, 100%

100% of nozzles nozzles were inspected byvisual by visual examination.examination.

Since the video was void of head orientation narration, orientation narration, each specific nozzle nozzle view could not be correlated by nozzle nozzle number.

Nozzles Nozzles 1,2,3, 1,2,3, and 4 which do not have sufficient interference interference gap were were excluded. excluded.

The remaining 65 nozzles nozzles did not show any evidence evidence of leakage.

leakage.

Bold Bold letters indicate indicate leaking CRDM bolting bolting flanges discovered discovered and repairedrepaired during 12 12 RFO ( April April 2000).

No No Leak Observed = Visual Inspection Satisfactory, Observed = Visual Inspection Satisfactory, No Video Record Required.

=

No Leak Recorded = NozzleNozzle inspection inspection recorded on videotape videotape Italicized Italicized text indicates indicates nozzles nozzles that are not expected expected to show leakage due to insufficient show leakage insufficient gap.

• FENOC RESTRICTED FENOC RESTRICTED INFORMATION INFORMATION

RPI!

RPYVHfd Hllad 11 115& & 12 121FFORIO InsDllclion inspection RlIsults Results

@ @ E>

0 @

Affected area Affected area since 11 RFO from leaking from leaking flange(s) flange(s) a - No leakage identified

@) identified o -

0 Evaluated not to have sufficient gap to exhibit leakage leakage

• Insufficient gap with leaking

- Insufficient leaking flange

• o -

0 Nozzle obscured by boron

- Nozzle obscured by boron with leaking flange

• - Newly affected, since 11 RFO, by leaking flange(s)

FENOC RESTRICTED RESTRICTED INFORMATION

Spring 1996 Spring 1996 Inspection Inspection

,y FENOC RESTR FENOC ICTED INFORM RESTRICTED ATION INFORMATION

1996 Inspections 1996 Inspections The following pictures are representative representative of the head in the Spring 1996 Outage. The head was relatively relatively clean and afforded afforded a generally generally good inspection.

FENOC RESTRICTED INFORMATION FENOC C 0"2-C. 0--

1996 Inspections FENOC RESTRICTED INFORMATION

1996 rInspections 1996 nspections Some boron piles were observed observed at the top of the head in the vicinity of previous leaking flanges. Because of its of its location on the location on the head, head, it it could could not be removed by mechanical mechanical cleaning but was verified to not be active or wet or and therefore wet and therefore did not pose did not pose aa threat threat to the the head from a corrosion corrosion standpoint. Additionally, since these drives are drives are not credited with not credited with leaking, leaking, that further ratifies that the boron is from previous flange leakage. The boron was heaviest beneath the mirror insulation seams.

FENOC RESTRICTED FENOC RESTRICTED INFORMATION INFORMATION

1996 Inspections 1996 Inspections FENOC FENoe RESTRICTED RESTRICTED INFORMATION INFORMATION CQS CuE;

1996 Inspections FENOC RESTRICTED FENOC RESTRICTED INFORMATION (7? (j (00

1996 Inspections 1996 Inspections FENOC FENOC RESTRICTED RESTRICTED INFORMATION INFORMATION C07 COr

1996 Inspections Inspections FENOC RESTRICTED RESTRICTED INFORMATION CQ~

1996 1996 Inspections Inspections Hole 44-45 FENOC RESTRICTED INFORMATION INFORMATION CoC{

1996 Inspections 1996 Inspections Hole 37-38 37-38 FENOC RESTRICTED INFORMATION INFORMATION C- ý

1996 Inspections 1996 1W,1ý 33-34 Hole 11-1 FENOC RESTRICTED FENOC RESTRICTED INFORMATION INFORMATION Cl!

1996 Inspections FENOC RESTRICTED INFORMATION INFORMATION

\-z.

C I-?

1996 Inspections 1996 29-30 Hole 29-30 The The boron deposits uphill of the CRDM drive below boron deposits below and to the right was reviewed reviewed from several angles and definite definite trails trails of of born could be seen streaming from above the mirror insulation. This coupled born could coupled with no no boron on the bottom (downhill) edge of the CRDM penetration the CRDM penetration and the fact that boron will grow but not flow uphill allowed allowed us to call this penetration penetration as a non-leaker.

FENOC RESTRICTED FENOC RESTRICTED INFORMATION INFORMATION CK2

"\

,v Spring 1998 Inspection Inspection F~OC FENOC RESTRICTED INfORMATION INFORMATION

RPI!

RPYIOuadI Hllad 11RIO 1FO IOS/lBClillo InspectANio RlIsuhs eIsuls

@> e @ @

e @ @ @ @

@ @ 0 @ @

@)

e @ 0

@) @

@ 28 Affected area from 1eaking flange(s)

()

@ -- No No leakage leakage identified identified Oo --Evaluated Evaluated not not to to have have sufficient sufficient gap gap to to exhibit exhibit leakage leakage

• -- Insufficient Insufficient gap gap with leaking flange with leaking flange o -

0 - Nozzle Nozzle obscured obscured byby boron boron

• -- Nozzle Nozzle obscured obscured byby boron boron with leaking flange with leaking flange FENOC RESTRICTED FENOC RESTRICTED INFORMATION INFORMATION

(' /4 ILl-

No.53 No.53 The following pictures are from access hole #9. They were clipped from from video taken in the Spring 1998.

Spring of 1998.

Although much more boron dusting was present in J1998 1996, a 998 than in 1996, good video inspection was able to be be performed for those 50 drives that obscured by boron from were not obscured leaking CRDM CRDM flanges.

flanges . Although much more video can be viewed, these these attached pictures representative of pictures are representative of condition of the drives and the the condition the heads. We attempted to capture in still photographs all of the outer most drives since they are the most susceptible circumferential cracking susceptible to circumferential cracking element analysis based upon finite element analysis showed them to have the highest which showed stresses on the uphill and downhill stresses slopes of the penetration.

penetration.

NO. 65 NO. What can also be seen in many of the the photos is the staining of the underside underside of the mirror insulation by boron trails.

This corresponds corresponds to the boron found on insulation in the top of the mirror insulation the vicinity of the vicinity leaking CRDM the leaking CRDM flanges.

flanges.

FENCC RESTRICTED INFORMATION FENOC RESTRICTED INFORMATION c,

CL 5"'

15

NO.

NO. 41 No. 33 33 No.48 RESTRICTED INFORMATION FENOC RESTRICTED INFORMATION cl(p

No. 65 No. 65 FENOC RESTRICTED INFORMATION INFORMATION (I?

The The two pictures pictures to the the left left are examples examples of some some drives where where we had had to view them from several several angles to ascertain ascertain that that the boron boron adjacent adjacent to the drives was drives was actually actually boron boron that flowed or tumbled down from higher higher up on the head and head and came to rest against the uphill uphill side of the the CRDM nozzle.

nozzle. Sometimes Sometimes this was was ascertained ascertained by comparing comparing thethe pictures pictures atat the left to video of the vacuuming that that was performed performed later later which showed the the boron to very loose and not a crystalline crystalline mass.. Additionally, mass Additionally , there were no boron boron deposits on the the downhill penetration seam, which is contrary, contrary , to what industry Industry experience experience has shown us to be true true at leakers.

plants that have identified leakers.

Because Because of the tight tolerances of the the penetrations, leakage through the penetrations, any leakage penetration penetration will encircle encircle the drive with the largest accumulation accumulation being on the the downhill edge because because of gravity flflowow to location.

that location.

FENOC RESTRICTED INFORMATION INFORMATION c,~

No. 62 FENOC RESTRICTED RESTRICTED INFORMATION INFORMATION

Note the Note the loose loose boron boron clumps clumps to to the the left which were not in in the the immediate immediate vicinity of vicinity of the nozzle penetrations.

the nozzle penetrations.

clumps appeared to These clumps to have have accumulated further accumulated on the up on further up the head head and then rolled or tumbled to and then rolled or tumbled to their their spots as resting spots resting as shown. Note also shown. Note the also the boron traces around the mirror insulation penetrations.

insulation penetrations.

No. 50

)U FENOC FENOC RESTRICTED RESTRICTED INFORMATION INFORMATION CZ-<;J

No. 63 No.

NO .35 40-.13 No.

NO. 42 4z FENOC RESTRICTED INFORMATION cC z..'

Z- ý

No.

iN(). 13 1.

~4i INO. 43 No.

No. 60 FENOC RESTR ICTED INFORMATION RESTRICTED INFO&~TION cz..

c.Z7ý7.

No. 24 No.43 No. 67 67 FENOC RESTRICTED INFORMATION FENOC RESTRICTED INFORMATION C23

No. 48,54,66 48, 54, 66 No. 67 FENOC RESTRICTED INFORMATION C2?ZL-

Penetrations as viewed from CRDM Penetrations inspection opening inspection opening #7 No. 56 No. 29 29 No49 side RESTRICTED INFORMATION FENOC RESTRICTED

('ZS

No. 55 No. 49 tront No. front No.36 No .36 FENOC RESTRICTED INFORMATION FENOC RESTRICTED 7 f CZt

No. 68 No. 44 No. 61 FENOC RESTRICTED RESTRICTED INFORMATION C -Z-ý

No. 25 tar.ir I

No. 61 INO. 01 No. 25 FENOC INFORMATION FENOC RESTRICTED INFORMATION C2~

No.

No. 68 No. 69 iNo. and No.

05, ana iNo. 4) tne mcicile 45 in the middle on tfle the back Dack FENOC RESTRICTED FENOC RESTRICTED INFORMATION INFORMATION (A

No.

NO. 57 No. 4b NO. 46 No. 57 No. 57 FENOC RESTRICTED INFORMATION INFORMATION

(<A>2 ý

NO. 37 NO. 3?

No. 26 No. 48 FENOC RESTRICTED INFORMATION INFORMATION

No. 34 11,0. 34-Same as above No. 34 on the right No. 2828 FENOC RESTRICTED INFORMATION INFORMATION

No. 48 No. 48 No. 66 No.

No. 18 FENOC RESTRICTED INFORMATION

(

No.

NJo. 59 Zf No. b9 NO. 59 No. bTI NO. 52

)

FENOC RESTRICTED FENoe RESTRICTED INFORMATION INFORMATION c- ý3qL

No.

IN0. 59 WD FENOC RESTRICTED RESTRICTED INFORMATION

Spring 2000 Inspection Inspection FENOC RESTRICTED RESTRICTED INFORMATION

IPVIead R" Iliad 12 121Ff? IiNSiuCl Rllsulls RID IOBl/llelioB lusuls

@ @ @)

@) e @)

@) @ @ @

@) @ 0 @ @ @

@) 0

@)

0

• Q e

Affected area from leaking flange(s) identified

@ - No leakage identified o -

O - Evaluated Evaluated not to have sufficient sufficient gap to exhibit leakage V . - Insufficient

. Insufficient gap with leaking flange O - Nozzle

.. Nozzle obscured by boron

• - Nozzle obscured by boron with leaking leakingflange flange RESTRICTED INFORMATION FENOC RESTRICTED

These photos were taken from our 2000 These spring spring outage videotapes.

The lighting and video video camera optics created an orange coloration of all of the created However, deposits pictures. However.

pictures. deposits of boron are visually discernable discernable as shown by the the scattered pieces of boron.

boron.

No 67 has no buildup buildup around its penetration penetration and the boron debris shown in the picture picture for No. are scattered well away from the No. 43 are the penetration.

No. 67 67 These drives These drives were were video taped because they taped because had boron deposits in the vicinity of the the CRDMs. Completely clean drive drive penetrations are not depicted here.

penetrations The photo for No No.. 19 depicts in the the background the extent of boron buildup on background on the head and is the reason no credit is taken taken for being able to visually inspect the the remainder remainder of the drives.

drives.

No. 43 No. 35 3J-FENOC RESTRICTED INFORMATION FENOC

No.

No. 60

)

No. 24 The debris piled up against the uphill side of No. 66 on the next page is is indicative of loose debris that has fallen down the slope of the head and drive. It does not came to rest on the drive. not resemble "popcorn" deposits witnessed resemble "popcorn" witnessed at other plants. There were also no no signs of boron anywhere else on the drive penetration opening.

penetration opening.

No. 19 FENOC RESTRICTED INFORMATION 41

No. 66 No. 66 No.

No.

No. 42 No. 19 No. No. 24 No.

No.

No. 35 35 35 No. 35 FENOC RESTRICTED FENOC RESTRICTED INFORMATION INFORMATION

No. 55 No. 29 FENOC FENOC RESTRICTED INFORMATION (744-<

Docket Number 50-346 Docket 50-346 License License Number Number NPF-3 Serial Serial Number Number 2744 2744 Attachment Page Page 1 of of I1 10 10 CFR CPR 2.790 Affidavit Affidavit (2 pages pages follow)

AFFIDAVIT OF STEVEN STEVEN P. MOFFITT MOFFITT A. My name is Steven P. Moffitt. I am Director Director - Technical Services Services for FirstEnergy Nuclear Operating Company ("FENOC") at the Davis-BesseDavis-Besse Nuclear Nuclear Power Power Station, Unit 1 ("DBNPS-l

("DBNPS-1"), "), and as such, I am authorized authorized to execute this Affidavit.

B. I am familiar with the criteria applied by FENOC to determine whether certain FENOC FENOC information is proprietary proprietary and I am familiar with the procedures procedures established with with FENOC FENOC to ensure the proper application application of these criteria.

C. I am familiar with the FENOC information information included in the DBNPS-l DBNPS-1 Tenth Refueling Refueling Outage, Eleventh Eleventh Refueling Outage, and Twelfth Refueling Outage Reactor Reactor Vessel Head Inspection Inspection photographs photographs and hereto referred to as "Photographs". Information contained in these Photographs has been classified Information contained by FENOC as Restricted Restricted in accordance accordance with the policies established by FENOC FENOC protection of confidential and proprietary for the control and protection proprietary information.

D. These Photographs Photographs are being made available available to the U.S. Nuclear Regulatory Commission Commission in confidence confidence with a statement that it is Restricted information information and a request that the information contained contained in these Photographs Photographs be withheld withheld from public disclosure.

E. The following information is provided to demonstrate that the provisions provisions of the Federal Regulations, Title 10 - Energy, Part 2, Section 790 have been Code of Federal considered considered in the confidential and commercial commercial classification of these Photographs as Restricted:

(i) These Photographs have been held in confidence confidence by FENOC. Copies of these Photographs are clearly marked marked as Restricted.

(ii) These Photographs Photographs contain information proprietary and confidential information of a proprietary nature and is of the type customarily customarily held in confidence confidence by FENOC and not made available available to the public.

(iii)

(iii) These Photographs Photographs are being being transmitted to the U.S. Nuclear Regulatory Commission Commission in confidence.

(iv) . These Photographs are not available available in public sources.

sources.

(v) These Photographs contain contain confidential and commercial information regarding regarding the material condition condition of certain components of the DBNPS-1 DBNPS-l that can be subject subject to future negotiated commercial commercial purchase agreements agreements with a vendor(s) vendor(s) external to FENOC. The information providedprovided on these Photographs is of a nature that cannot be acquired or duplicated by others.

F. In accordance accordance with FENOC's FENOC's policies governing governing the protection and control ofof information, Restricted inforni.ation, Restricted information information contained contained in these Photographs has been made available, on a limited basis, outside FENOC only as required and under suitable non- disclosure agreement agreement providing providing limited use of the information.

I

G. FENOC requires that Restricted information FENOC contained in these Photographs infonnation contained Photographs be be secured file or area and distributed only on a need-to-know kept in a secured need-to-know basis.

H. The foregoing statements statements are true and correct to the best of my knowledge, knowledge; infonnation, information, and belief.

By:A~@1?Y~

By:

Steven P. Moffitt Affinned and subscribed before me this 30th Affirmed October, 2001.

30th Day of October, 2001.

4, .. 1?~

Notary Public~

Public, State of Ohio My Commission Expires August 16, 16, 2006

)

22

Docket Number 50-346 50-346 License License Number NPF-3 Serial Number 2744 2744 Attachment 3 Attachment Page 1I of I1 Commitment List List The following list identifies those actions committed committed to by the Davis-Besse Nuclear Power Power Station (DBNPS) in this document. Any other actions discussed in the submittal represent represent intended or planned planned actions the DBNPS. They are described described only for information and are not regulatory commitments. Please notify the Manager information Manager - Regulatory Affairs (419-321-8450) at the DBNPS of any questions questions regarding regarding this document or associated associated regulatory commitments.

commitments.

COMMITMENTS COMMITMENTS DUE DATE None

)/"'"

NRC LETTER,S*

.NRC LETTERS - REVIEW AND APPROVAL APPROVAL REPORT ED 7159-7 U.S. NRC NRC "'.{- {.~ {LI

() ~C /' _ (. S./ .----- In f9re DAVID DAVID GEISEN B'0 wIo#mo

, .f,

.Exh!bIt'-l_

fl n ~ '-- ~ ~ Docket # 1A-05-052 1A-05-052 .

(3)

SUMMARY

1(3) No** TitlB

SUMMARY

(Log No.. Subject)

Tide Subject) . . . . . Date Marked ID:tI.., 2008 Marked for IO:l1d.L. 2008 (Tr. p. gFZ25 . ))

,. ,;> -~STransmittal Transmittal of Results .of of Results of RVP RVP Head Head CRDM CRDM Nozzle Nozzle Visual Examinations Visual Examinations .. E' I ~I i 2008 (T g2,& )

. COMMITMENT rmiPkn~n-rremT UST IST~ ADDED Anntl~n TO ToA iLETTER mrIrh !~i 15) PERIODI Offered In Date Offered in Ev: iz.(2, 2008 (Tr.r. pp...oo:_____

V . ..L..!:::+.2-, gtb -I_

IA:rERiEsPc5NsE~~BESUBMmeO;:O~:-----17liWiYECES:iALi~

t r) I DATE RESPONSE DUE TO BE SUBMITTED TO NRC NRC I.

I Through Wi~es Through Witnes nel: N/?<- -

Target Date. 11/01/01 11/01/01 EXPRE Action:

Action:. A MITI ITT REJECTED WITHDRAWN REJECTED WITHDRAWN F1 WA 10 EXPRE

~~~~~.;.;..:;.-=----------.a....:..:.:...:.--t(~9)~NOTmiiAR'iVy Date: (Tr. p. 82(;) _

(8) PREPARED BY Date; i1{..L..2008

('a 2 00 (TrS .SE 8

Rod Cook ext. 7782 ext. 7782 181 YES [ . .

~~~~~~=-~--------------~~~~--~~~~

(11) ADDITIONAL REFERENCES (12) ~UMMITMENT NO.(S) CLOSED uymmrUMENT NO.(S) CLOSED Serial 2731 Serial 2735 DOCKETED USNRC USNRC 1-------=---------,----------------'---------....-

(13) COMMENTS September 9, September 2009 (11:00am) 9,2009 (11:00am)

OFFICE OF OFFICE OF SECRETARY SECRETARY RULEMAKINGS AND RULEMAKINGS AND ADJUDICATIONS ADJUDICATIONS STAFF STAFF (14) REVIEW AND APPROVAL REVIEW 0181 COGNIZANT COGNIZANT REGULATORY REGULATORY AFFAIRS INDIVIDUAL INDIVIDUAL R.M.

R.M. Cook 0181 MANAGER, MANAGER, DESIGN ENGINEERING DESIGN ENGINEERING D. Geisen Geisen 0181 DIRECTOR, TECHNICAL TECHNICAL SERVICES SERVICES . S. Moffitt 0o o

o 0

o o

181 04 SUPERVISOR, SUPERVISOR, 08 DB UCENSING UCENSING D.R. Wuokko 181 0 SUPERVISOR.

SUPERVISOR, 08 D0 COMPUANCE COMPLIANCE D.L.

D.L. Miller 0181 MANAGER, REGULATORY AFFAIRS MANAGER, REGULATORY D.. H. Lockwood D. Lockwood 181 0 VICE PRESIDENT A1 G.G. Campbell OCT 30 2001 (17) AOOmONAL DISTRIBlJTlON Sl1-00602 511-00602 NRC027 -1702 NRC027-1702

NRC LETTEHS

.NRC LETTER.S -* REVIEW AND APPROVAL REPORT ED 7159-7 7159-7 (1) ~RDS (1)

~_C;~O MANAGEME~,~0rl P!q*O0RDS MANAGEMENT

-l~*_;- l-* N 7a

/ ....00..,/'V'_7 \2744 (2) SERIAl 2744 1.'(2) SERIAL NO.

NO.

(3).

SUMMARY

(3)

SUMMARY

(Log No., TlUeSubject)

(Log No.. TitleSubject)o .... -* " -

Transmittal of Results Transmittal Results of RVP Head CRDM Nozzle Nozzle Visual Examinations Examinations (4) COMMITMENT uST ADDED TO LETTER (5) PERIODIC / NON-PERIODIC REPORT

• (4) COMMITMENT UST ADDED TO lETTER 181 (5) PERIODIC I NON~PERIODIC REPORT 0 YES DYES 0181 NO NO REPORT REPORT NO. NO.

(6) DATE RESPONSE (6) RESPONSE DUE TO BE SUBMITTED SUBMITTED TO TO NRC 7) 7} SPECIAL J/tC. £TI'tSA-N/J!3 HANDLING R/,41 It SPECIALHANDUNG TT,4IV_ IDATE DATE SENT SENT Target Date 11/01/01 []N/A ON/A 0 EXPRESS DELIVERY 181 DELIVERY 01TELECOPY0 TELECOPY I 10 ..30-0 I

/O-3-o (8) PR~ARED (8) PREPARED BY NOTARY (9) NOTARY (10) LICENSE FEE REQUIRED 1(10) REQUIRED Rod Cook ext. 7782 181 OYES YES ONO D NO 10YES DVES ONO ~ NO (11) ADDITIONAl. REFERENCES (11) ADDITIONAL REFERENCES (12) COMMITMENT NO.(S) CLOSED (12) COMMITMENT Serial 2731 Serial 2735 (13) COMMENTS COMMENTS (14) REVIEW AND REVIEW APPROVAL AND APPROVAL INITIALS INITIALS DATE

.DATE RECEIVED f::II=r.I=IVI=D. APPROVED APPROVED 1810COGNIZANT COGNIZANT REGULATORY REGULATORY AFFAIRS AFFAIRS INDIVIDUAL INDIVIDUAL A.M. Cook R.M.

~A "'Iso k( 10 12co/2t?D (f b? Cv

~/?1)a>1

• 0 MANAGER, DESIGN ENGINEERING 181 ENGINEERING

• o,. o..,*ICc..Lc.vc D. Geisen D.

~. IO/JO/z.co

, , I /6.,o,,0 Zo*.

16/30/ P .PI z.::o I -

I------,...,---~~~~~ , /

1810DIRECTOR, DIRECTOR, TECHNICAL TECHNICAL SERVICES SERVICES S. Moffitt S.Moffitt

~~~

o o

o 0

o o

0181 SUPERVISOR, 08 DB UCENSING UICENSING D.R. Wuokko D.A. Wuok ko 14 I c[01]Qj-- . leila/or 181 0 SUPERVISOR, SUPERVISOR, 08 DB3 COMPUANCE COMPUIANCE D.L. Miller _______

[tJlc~lc>1 1810MANAGER, MANAGER. REGULATORY REGULATORY AFFAIRS AFFAIRS D. H.

D. H. Lockwood

-~

'" ,.£2- /

0181 VICE PRESIDENT VICE PRESIDENT /l G.G.camp~II' G.G. Campbell 'OT3020 ~} OCT 30 2001 DATE ADDED TO LETtER DATE ADDED LET-TER &;I'l ~(15) ADDEDBY (17) ADDITIONAL (17) ADDITIONAL DISTRIBUTION DISTRIBUTION T

/0 "'3d-tJL DATE SENT j~

DATE SENT TO NRC NRC 6)TRIBUTED BY

\" ~BUTED BY /J ..

,P-3I-01 '" '-'1.1 (~."'" ....

• A -

DATE OF ILIND D!VrTIBUTION (1 STRIBUTEDB I .

VI Sll-00602 511*00602 NRC027-1702 NRC027-1702

Approval Report (ED 71~9-7)

Review and Approval The NRC Letters --Rev.lew 7159-7) should be completed completed by.the Regulatory Affairs Section by the Regulatory Section..

.. ~ .

BLOCK 1 MANAGEMENT NO.-

RECORDS MANAGEMENT NO. - Regulatory Regulatory Affairs enters Records Management number prior tcW Records Management distribution correspondence to NRC. .

distribution of correspondence W BLOCK 2 SERIAL NO.

SERIAL. NO. -Initiator"enters obtained from the Regulatory Affairs Clerk.

number obtained

- Initiatbrenters serial number Clerk.

BLOCK 3

SUMMARY

(Log No.,

SUMMARY

correspondence. This summary No., Title Subject) - Initiator enters a summary of the correspondence.

should identify ififthe correspondence correspondence is in in response to any previous correspondence and why the correspondence the letter is letter is being written.

BLOCK 4 COMMITMENT COMMITMENT USTADDED LIST ADDED TO TO LETTER - Preparers block to indiate a commitment Preparers checks the block commitment list has included with the letter.

been included

.been BLOCK 5 BLOCKS PERIODIC/NON-PERIODIC REPORT PERIODIC/NON-PERIODIC REPORT :Identify correspondence is a Periodic or Non-

-'Identify whether this correspondence Periodic Report as identified Periodic identified in Nuclear Group Procedure NG-NS-00807.

BLOCK 6 RESPONSE DUE TO BE SUBMITTED DATE RESPONSE SUBMITTED TO NRC - Initiator correspondence is Initiator enters the date the correspondence NRC. IfIfthe correspondence due to the NRC. correspondence does not have a required required due date, the block block shall be marked not applicable (NA).

BLOCK 7 HANDLING - Initiator checks if SPECIAL HANDLING correspondence requires special distribution to the NRC. If if the correspondence If yes, the Regulatory Affairs clerk enters date the correspondence sent.

correspondence is .sent.

• BLOCK BLOCK 8 PREPARED BY -Initiator PREPARED names of individuals responsible for providing

- Initiator enters the names providing technical information information correspondence along with hislher for the correspondence name.

his/her name.' ... .

BLOCK 9 NOTARY correspondence is required to be notarized.

NOTARY - Initiator checks ififthe correspondence to BLOCK 10 LICENSE FEE REQUIRED - Initiator checks ififa license fee is required, LICENSE required, per the requirements of 11 Part 170. If initiator shall complete a Voucher If yes, the Initiator Voucher Check Authorization (Form 294) and obtainO appropriate fees to accompany correspondence.

accompany the correspondence.

BLOCK 11 REFERENCES - Initiator ADDITIONAL REFERENCES Initiator enters any additional NRC correspondence correspondence or documents documents that pertain to the subject correspondence. .

BLOCK 12 COMMITMENT NO(S). CLOSED COMMITMENT CLOSED - Initiator enters the Commitment Management Management System number(s) of any commitments that are closed by the subject ariy correspondence.

subject correspondence.

BLOCK 13 COMMENTS - Jnitiator COMMENTS ~nitiator or any reviewer enters appropriate comments regarding the subject enters appropriate subject correspondence.

correspondence.

BLOCK 14 APPROVAL*- Initiator checks REVIEW AND APPROVAL REVIEW checks and lor /or enters reviewer(s). The technical enters the desired reviewer(s).

accuracy of a response to the NRC is the responsibility accuracy Management individual responsibility of the Director and Management assigned assigned the action. .

BLOCK 15 DATE ADDED BY - Distributor checks the Date Added to Letter Block and signs the Date Added By By block indicate the original letter has been dated prior to distribution to the NRC.

blpck to indicate BLOCK 16 Distribution to the NRC DISTRIBUTED BY - Distribution DISTRIBUTED NRC shall be made made by the Regulatory Affairs $ection. Distributor Section. Distributor signs the Distributed By block and completes the Date Sent to NRC block.

BLOCK 17 DISTRIBUTION - Initiator enters ADDITIONAL DISTRIBUTION ADDITIONAL individuals requiring enters individuals not on the requiring distribution that are riot the standard distribution standard list.

distribution list.' . . .

BLOCK 18 DISTRIBUTED BY - Distributor signs the Distributed By'blqck DISTRIBUTED By block and completes the Date 6f Blind" 6f"'Blind' Distribution block.

si1-00603 511-00603 NRC027-1703 NRC027*1703

1ý + -,S> C-i ~ U.S. NRC In NRC GEISEN ;nJAl~

(. L (1{'

Exhibit#_~

Exhii 15

1. t1 --

in re DAVID GEISEN DOCKETED USNRC Docket # 1I A-05-Q52 A-05-052 .' .5 USNRC Marked for 10:11:/.1-.2008 Date Marked ID'I.L, 2008 (Tr. P.-~

September 9, 2009 (1 1:00am)

September 9, 2009 (11 :OOam)

OFFICE Date Offered in EV:..lifi-.

in Ev: . ...8. L-L) 2008 (Tr. pp.2 2, G )

OFFICE OF OF SECRETARy SECRET Through Witness/Panel:_~N~/.:..pr / f/_ _ _ _ __

RULEMAKINGS RULEMAKINGS AN~RY AND Through Witness/Panel:_'

ADJUDICATIONS STAFF ADJUDICATIONS STAFF ActiO; ~

Action: M WITHDRAWN REJECTED WITHDRAWN Date: ~0084 2Z tr. (Tr. P.3 '2 (& )

§ 50.5

§50.5 10 CFR Ch. II (1-1-08 Edition)

§50.55(f)(3). or a change

§50.55(f)(3), change to a licensee's licensee's (e) Conflicting (e) requirements. The com-Conflicting requirements.

NRC-accepted quality assurance assurance top- requirements contained munications requirements contained in ical report under § 50.54(a)(3) 50.54(a) (3) or this section and §§ §§50.12, 50.30, 50.36, 50, 12. 50,30, 50.36,

§ 50.55(1) (3). must be submitted to the

§50.55(f)(3). 50,36a. 50,44.

50.36a. 50.44, 50.49.

50.49, 50.54, 50.54, 50.55.

50.55, 50.55a, 50.55a.

NRC's Document Control Desk, with a 50.59, 50.62, 50.59. 50.62, 50.71.

50.71. 50.73, 50.82, *50.90.

50,73. 50.B2. :50.90, and copy to the appropriate appropriate Regional Of- 50.91 supersede and replace replace all existing fice, and a copy to the appropriate appropriate NRC requirements in any license conditions requirements Resident Resident Inspector if If one has been as- or technical specifications speCifications in effect effect on signed to the site of the facility. If If the January 5. 1987. Exceptions to these re-5, 19B7, communication communication is Is on paper.

paper, the sub- quirements must be approved quirements approved by the mission to the Document Document Control Desk Office of Information Information Services. Nuclear Nuclear must be the signed original. Regulatory Commission, Washington, Regulatory CommiSSion, Washington.

(ii) A change to an NRC-accepted (ii) NRC-accepted 20555-0001, -telephone (301)

DC 20555-0001, 415-7233, (301) 415-7233.

quality assurance assurance topicaltopical report from e-mail INFOCOLLECTS@nrc.gov.

e-mail INFOCOLLECTS@nrc.gov, nonllcensees (I.e" nonlicensees architect/engineers, (i.e,, architect/engineers, 168FR 58808.

168 58808, Oct. 10, 20031

10. 2003J suppliers. fuel suppliers, NSSS suppliers, suppliers. con-structors.

structors, etc,)"etc.) must be submittedsubmitted to to §50.5 Deliberate misconduct.

§ 50.5 Deliberate the NRC's Document-Control Document. Control Desk, If If (a) Any licensee, licensee, applicant for a li- li-the communication communication is on paper, paper. the cense, employee employee of a licensee licensee or appli-signed original must must be sent. cant; cant; or any contractorcontractor (including a CertlfJcatioTJ of permanent

, (8) Certification permanent cessation cessatIon supplier supplier or consultant),

consultant), subcontractor.

subcontractor, of operations.

operations. The licensee's licensee's certifi- employee employee of a contractor or subcon-cation of permanent cation cessation of oper-permanent cessation tractor of any licenseelicensee or applicant for for ations.

ations, under §§50.82(a)(1),

50.82(a)(l). must must state a license, who knowingly provides to the date on which operations operations have licensee, applicant, contractor.

any licensee. contractor, or or ceased ceased or will cease, cease. and must be be sub- subcontractor, any components, equip-mitted to the NRC's Document mitted Document Control ment, materials, ment. materials. or other goods or or Desk. This sub.mlssion submission must be under under services that relate to a licensee's licensee's or oath or affirmation.

affirmation. applicant's activities activities in this part. part, may (9) Certification permanent fuel re-Certification of permanent re- not:

moval.

moval. The IIcensee's certification of licensee's certification of l() Engage in (1) In deliberate misconduct misconduct permanent permanent . fuel removal.

removal, . under under that causes causes .or or would have caused. caused, if

§ 50.82(a)(1). must state

§50.82(a)(1), state the date on not detected, a licensee licensee or applicant applicant to which which the fuel was removed from the be inIn violation violation of. of any rule, regulation, reactor vessel and the disposition disposition of of or order; order; or any term, term. condition, condition. or or the fuel, fuel. and must be submitted submitted to the limitation limitation of any license issued by the thc NRC's Document Control NRC's Control Desk. This Commission; or or submission must be under oath or affir-submission (2) Deliberately submit to the NRC, a (2) Deliberately mation. licensee, licensee. an applicant, or a licensee's Iicensee's (c) Form Form of communications.

communIcatIons. All All paper paper or applicant's contractorcontractor or subcon-copies copies submitted to meet the require- tractor, Information tractor. information that the person ments set forth In in paragraph paragraph (b) of this submitting submitting the informationinformation knows knows to section must be typewritten, section typewritten, printed or or' incomplete or inaccurate be incomplete inaccurate in some otherwise otherwise reproduced reproduced in permanent permanent respect respect material to the NRC.

form on unglazed unglazed paper. paper, Exceptions Exceptions to to (b) A person who violates paragraph paragraph these these requirements imposed Imposed on paperpaper (a)(])

(a)(I) or (a)(2) of this section may be submissions may be granted for the the subject subject to enforcement enforcement action action in ac-submission submission of' of 'micrographic, micrographic, photo- cordance with the procedures cordance procedures in 10 CFR graphic, or similar similar forms.

forms, part 2, 2. subpart B. B, (d) Regulation governIng submission, Regulation governing submission. (c) For the purposes of paragraph Licensees and Licensees and applicants applicants submitting (a)([)

(a)(I) of this section, deliberate deliberate mis-correspondence.

correspondence, reports, and other other conduct conduct by a person means means an inten-written communications communications under under the reg- tional tional act or omission that the person ulations ulations of this part are requested requested but but knows:

not required reqUired to cite whenever whenever prac- (1) Would cause a licensee or appli-(1) appli-tical, in the upper tical. upper right corner corner of the cant to be in violation violation of any rule, rule. reg-first page of the submission,submission. the spe- ulation, or order; or any term. term, condl*

condi cific regulation or other other basis requiring tion, or limitation, of any license submission. issued by the Commission; Commission: or or 728 728

-72rý,,. ý,L. r6-- E,ýC y C,

Nuclear Regulatory Nuclear Regulatory Commission §50.7

§ 50.7 (2) Constitutes a violation of aa re-(2) prohibited by this section who, acting acting quirement, procedure, instruction, con- without direction direction from his or her em-purchase order, or policy of a li-tract, purchase ployer (or the employer's employer's agent).

agent), delib-censee, applicant, contractor, or sub- erately causes causes a violation violation of any re-contractor. quirement quirement of the Energy Energy Reorganiza-163 FR 1897, 163 1897, Jan. 13, 19981 1998] tion Act of 1974, 1974, as amended, or the Atomic Energy Energy Act of 1954, 1954, as amend-

§50.7 Employee protection. ed.

(a) Discrimination (a) Discrimination by a Commission Commission (b) Any employee who believes that that licensee, an applicant licensee, applicant for a Commis- he or she has been discharged or other-sion license, or a contractor or subcon- wise wise discriminated against by any any per-Commission licensee tractor of a Commission licensee or ap- son for engaging in protected protected activities plicant against an employeeemployee for engag- specified in paragraph paragraph (a) (1) of this sec-(a) (I) ing in certain protected activities activities is is tion maymay seek a remedy for the dis-prohibited. Discrimination includes includes charge charge or discrimination discrimination through an discharge and and other actions that relate administrative administrative proceeding in the De-to compensation, compensation, terms, conditions, or or partment of Labor. The administrative privileges of employment. The pro-privileges .proceeding proceeding must be initiatedinitiated within 180 180 tected activities are are established established inin sec- 'days days after an alleged violation violation occurs.

tion 211 of the Energy Reorganization Reorganization The employee may do this by filing a The employee Act of 1974, 1974, as amended, and in generalgeneral complaint alleging alleging the violation violation with are related to the administration administration or en- the Department Department of Labor, Employment Employment forcement of a requirement requirement imposed imposed Standards Standards Administration, Wage and under the Atomic Atomic Energy Energy Act or the Energy Energy Reorganization Reorganization Act. Hour Division. The Department Department of of (1) The (I) The protected activities include include Labor may may order order reinstatement, back back but are not limited limited to: . compensatory damages.

pay, and compensatory damages.

Providing the Commission or his (i) Providing his (c) A violation of paragraph paragraph (a). (e),

(a), (e),

or her employer information' information' about al- al- or (f) of this section by a Commission leged leged violations of either of the stat- stat- licensee, an applicant applicant for a Commis-utes named named in paragraph paragraph (a) introduc- sion license, or a contractor contractor or subcon-tory text of this section section or possible possible vio- Commission licensee or ap-tractor of a Commission lations requirements* imposed under lations of requirements' under plicant may be groundsgrounds for- for-either either of those statutes; (1) Denial, (I) Denial, revocation, revocation, or suspension suspension (ii) Refusing Refusing to engage in in any any prac- of the license.

tice made unlawful under under either either of of the (2) Imposition Imposition of a civil penalty on on statutes named in paragraph (a) intro- the licensee, applicant, or aa contractorcontractor ductory text or under these require- or subcontractor subcontractor of the licensee licensee or ap-ments if if the employee has identified plicant. .

the alleged illegality illegality to the employer; (3) Other enforcement enforcement action.

(iii) Requesting the Commission (iii) Commission to to (d) Actions Actions taken taken by an employer, employer, or or institute institute action against his or her em-others, which adversely adversely affectaffect an em-ployer for the administration or Or en-requirem~ents; ployee may be predicated upon non-forcement of these requirements; (iv) Testifying in any Commission (iv) Commission discriminatory discriminatory grounds. The prohibi-proceeding, or before before Congress, or at* at tion applies applies when the adverse action action Federal or State proceeding proceeding re- occurs occurs because because the employee has en-any Federal gaged gaged in protected activities. An em-garding any provision (or proposed proposed pro-engagement in protected ac-ployee's engagement vision) of either of the statutesstatutes named in paragraph paragraph (a) introductory introductory text..

text. automatically render tivities does not automatically render (v) Assisting or participating participating in, or is htim him or her immune from discharge discharge or or about to assist or participate participate in, these discipline for legitimate discipline legitimate reasons or or activities. from adverse action dictated by non-(2) These activities are protected (2) prohibited considerations.

considerations.

.1 even even if if no formal proceeding is actu- (e)(1)

(e) (I) Each licensee licensee and each each appli-ally initiated as a result ally initiated the em-result of *the cant cant for a license license shall prominently prominently ployee ployee assistance or participation. post the revision of NRC Form 3, "No- 'No-(3) This section section has no application application to to tice tice to Employees," referenced in 10 Employees," referenced 10 any employee alleging discrimination alleging discrimination 19.11(c). This form must be posted CFR 19.11(c).

729 729

§ 50.8

§50.8 10 CFR Ch. I1 (1-1-08 (1-1-08 Edition)

Edition) sufficient to permit at locations sufficient permit em- (b) The approved (b) information collec-approved information collec-ployees protected by this section to ob-ployees tion requirements requirements contained in in this serve aa copy on the way:

serve way to, or or from part appear appear in §§ 50.30, 50.33, 50.34, in §§50.30, 50.34, their place of work. Premises must must be be 50.34a, 50.3'4a, 50.35, 50.35, 50.36, 50.36a, 50,36b,50.36b, 50.44, 50.44, posted not later posted later than 30 days after an an 50.46, 50.47, 50.48, 50.49, 50.46, 50.54, 50.55, 50.49, 50.54, 50.55, application is application docketed and, remain is docketed,and remain 50.59, 50.60; 50.61, 50.55a, 50.59, 50.55a, 50.62, 50.63, 50,61, 50.62, 50,63, posted while the application posted application is pending pending 50.64, 50.64, 50.65, 50.66, 50.68, 50.69, 50.70, 50.71, 50.65, 50.66, 50,71, before the Commission, during before during the 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 50.72,50.74,50.75,50.80,50.82,50.90,50.91, term of the license, and for 30 days fol-term appendices A, B, E, G.

50.120, and appendices G, H, I, lowing license termination.

termination, ], K, M, NO, J, N,O, Q, R, and S to this part.

Copies of NRC Form 33 may be ob-(2) Copies (c) This part contains information part contains tained by writing to the Regional Ad- collection collection requirements in addition to to ministrator of the appropriate appropriate U.S, U.S. Nu- approved under the control num-those approved clear Regulatory Regulatory Commission Regional Regional ber specified specified in in paragraph paragraph (a) (a) of this Office listed in in appendix D to part part 20 of of section. These information collection this chapter, by calling (301) (301) 415-5877, 415-5877, requirements and the control requirements control numbers forms@nrc.gov, or by vis-via e-mail to forms@nrc.gov, under which they are approved are as under iting the NRC's NRC's Web site' site at http://http:// follows:,,

follows; www.nrc.gov and and selecting forms .from selecting .forms, from 50.73, NRC Form (I) In §§50.73, (1) Form 366 is ap-index found on the home page.

the index proved under control number number 3150-0104.

3150-0104.

affecting the com-agreement affecting (f) No agreement (2)

(2) In §§50.78, 50.78, Form N-71 is is approved pensation, terms, conditions, or privi- under control control number 3150-0056.

leges of employment, including including an an agreement to settle agreement settle a complaint filed [49 FR 19627, May May 9, 9, 1984, 1984, as as amended amended at at 58 58 employee with the Department by an employee Department FR 68731, FR Dec. 29, 68731, Dec, 1993; 60 29, 1993; 60 FR 65468, Dec.

FR 65468, Dec, 19, 19, of Labor pursuant pursuant to section 211 of the 1995; 61 FR 65172, Dec. 11, 1996; 62 FR 52187, 1995; 52187, Reorganization Act Energy Reorganization Act of, 1974, as of'1974, Oct. 6, Oct, 6, 1997; 1997; 67 FR 67099, Nov. 4, 67099, Nov, 4, 2002; 2002; 68 FRFR 19727, Apr. 22.

19727, 2003; 69 FR 68046, Nov, 22, 2003;69 Nov. 22, 2004; amended, may contain any provision 70 FR 61887, Oct,Oct. 27. 20051

27. 2005J which would prohibit, restrict, or oth-erwise discourage discourage an employee employee from § 50.9 Completeness and accuracy Completeness accuracy of of participating participating in protectedprotected *activity activity as information.

information.

defined in in paragraph paragraph (a) (1) of this sec-(a)(1) sec-tion including, but not limited to, to, pro- (a) Information provided (a) Information provided to the Com-information to the NRC viding information NRC *or,-or , to to mission by an applicantapplicant for a license license or or employer on his or her employer potential viola-onpo'tential by .a information required a licensee or information tions or other matters matters -withinwithin NRC's NRC's by -statute Commission's reg-statute or by the Commission's regulatory responsibilities.

regulatory ulations, orders, or license conditions maintained by the applicant to be maintained applicant or or 158 FR 52410, Oct. 8,

[58 1993, as amended 8, 1993, amended at 60, FR

60. FR shall be complete and ac-licensee shall the licensee 24551, 24551, May 9, 9, 1995; 1995; 61 FR 6765, Feb. 22, 6765, Feb, 22, 1996; 1996; 6868 FR 58809, Oct. 10, 2003; ,72 Oct, 10.2003; 72 FR 63974, 63974, Nov. 14. 14, curate in all material respects.

2007]

2007J (b) Each applicant or licensee Each applicant licensee shall

§ 50.8 Information collection require- Commission of information notify the Commission information

§ 50.8 Information collection require- identified by the applicant identified applicant or or licensee ments: OMB approval., '

OMB approval. having for the regulated as having regulated activity a (a) The Nuclear Regulatory (a) Regulatory Commis- significant significant implication for public implication public submitted the information sion has submitted information health and safety or common common defense defense collection requirements contained collection requirements contained in in and security. An applicant applicant or licensee Office of Management part to the Office this part Management violates this paragraph only if if the ap-and Budget (OMB) for approval as' as, re- licensee fails to notify the plicant or licensee quired by the Paperwork Reduction Commission of information information that that thethe Act (44 U.S.C.

U.S,C. 3501 et seq.). seq.). TheThe NRC applicant or licensee 'has has identified identified as as conduct or sponsor, and a per-may not conduct implication for significant implication having a significant for son is not required to respond to, a col- public health and safety or common common de-lection information unless it lection of information it dis- fense and security. Notification Notification shall shall plays a currently currently valid OMB control be provided to the Administrator Administrator of the number. OMB has approved approved the infor- appropriate Regional Office appropriate Within two Office within mation requirements con-collection requirements mation collection working days of identifying identifying the infor-tained in in this part under control num- mation. This requirement is This requirement is not appli-ber 3150-0011.

3150-0011. information which cable to information which is already already I 730

Nuclear Regulatory Nuclear Regulatory Commission Commission §50.10

§50.10 required to be be provided provided to the Commis- mental mental mitigation mitigation measures, and and con-sion by sion by other reporting reporting or or updating updating re-re- struction struction of temporary temporary roads roads and and bor-bor-quirements.

quirements. row row areas; areas;

• (iv)

(iv) Erection Erection of of fences and other other ac-

[52

[52 FR 49372, 49372. Dec. 31, 31, 1987]

1987[

cess cess control control measures; REQUIREMENT REQUIREMENT OF OF LICENSE, LICENSE, EXCEPTIONS EXCEPTIONS (v) Excavation; Excavation; (vi) Erection Erection of of support support buildings

§§50.10 50.10 License License required; required; limited limited work work (such as, construction construction equipment equipment stor-authorization.

authorization. age sheds, warehouse warehouse and shop shop facili-Definitions. As used (a) Definitions. used inin this sec- ties, utilities, concrete concrete mixing plants, construction means tion, construction means the activities docking docking and and unloading unloading facilities, and in paragraph (a)(1) in paragraph (a) (I) of this section, section, and office office buildings) for use in connectionconnection does does not mean the activities in the activities in para-para- with the construction construction of the facility; graph graph (a) (2) of this this section. (vii) Building Building of serviceservice facilities, Activities constituting (I) Activities (1) constituting construc-construc- such such as pavedpaved roads, parking lots, rail- rail-tion are are the the driving driving of piles, subsurface subsurface road road spurs, exterior exterior utility and light-preparation, placement placement of backfill, ing systems, potable water systems, concrete, or permanent permanent retaining walls walls sanitary sewerage treatment sanitary sewerage treatment facilities, within an an excavation, excavation, installation installation of of and and transmission transmission lines; foundations, or in-place assembly, '(viii)

.(viii) Procurement Procurement or fabrication of of erection, fabrication, or testing, which which components components or portions of the proposed are for: facility facility occurring occurring at other other than than the final, final. in-place location location at the facility; (i)

(i) Safety-related Safety-related structures, sys-tems, or components components (SSCs) (SSCs) ofof aa facil- (ix) Manufacture Manufacture of a nuclear power power ity, as defined in 10 CFR CFR 50.2; reactor reactor under a manufacturing manufacturing license (ii) SSCs SSCs relied upon to mitigate mitigate ac- under under subpart F of part 52 of this chap-cidents or transients transients or used in plant plant ter to be installed at the proposed site emergency operating emergency operating procedures; and to be part of the proposed proposed facility; (iii) SSCs whose failure failure could could prevent prevent or or safety-related safety-related SSCs from fulfilling (x) With respect to production production or uti-safety-related function; their safety-related lization facilities, other than testing (iv) SSCs whose (iv) whose failure failure could cause a facilities and nuclear power plants, re-reactor scram or actuation of a safety- quired quired to be licensed under Section related system; 104.a 104.a or Section Section 104.c of the Act, the (v) SSCs necessary to comply with 10 (v) erection of buildings buildings which will be used CFR part 73; . for activities other than operation of a (vi) SSCs necessary to comply with 10 (vi) 10 facility and which may also be used to to CFR 50.48 and criterion 33 of 10 CFR part part house aa facility (e,g.,(e.g., the construction construction 50, appendix appendix A; and college laboratory of a college laboratory building with (vii) Onsite emergency facilities, that (vii) that space for installation of aa training re-is, technical support and operations actor).

actor),

support centers, necessary to comply comply (b) Requirement Requirement for license.

license, Except as with 10 CFR 50.47 and 10 CFR part 50, provided in §50.11§ 50.11 of this chapter, no appendix E. person within the United States shall (2) Construction (2) Construction does not include: transfer or receive in interstateinterstate com-(i) Changes Changes for temporary temporary use of the merce, manufacture, produce, transfer, land for public recreational purposes; acquire, possess, or use any productionproduction (ii) Site exploration, including nec- or utilization facility except as author- author-essary borings to determine determine foundation ized by a license issued by the Commis-conditions or other preconstruction sion.

sion.

monitoring to establish background background in- (c) Requirement (c) Requirement for construction constructionpermit, permit, formation related to the suitability suitability of of early site permit early permit authorizing authorizinglimited limited work the site, the the environmental environmental impacts impacts of of authorizationactivities, authorization activities, combined combined license, license, construction or or operation, or the pro- or limited or limited work authorization.

authorization.No personperson tection of environmental values; may begin the the construction construction of a pro-(iii) Preparation of (iii) of aa site. for con- utilization facility on a site duction or utilization site struction of aa facility,.

struction facility,, including clear-clear~ on which the facility is to be operated ing of the site, grading, installation of of until that person has been issued either either drainage, erosion and other environ- aa construction construction permit under this part, part, 731 731

Iý rf- 7Y__ VrL) I I

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0. SYSTEM CAPABLE CAPABLE OF PERFORMING PERFORMING SPECtFtED SPECIFIED FUNCTION 0o N NIA po~entlal/y reportable

..- Non Tech Spec and not a potentially reportable systBm ystem iJII YES ig - Provide Provlcfe' Justification unless Weighting .F~or Justification unl888 less than 11

.Fcor le88 o fII()

0 NO -- Initiale Initiate an an ad$ndum addendum ifif Part 2*1s Opqrable i .Yes 2 .1s O.,.rable Yas JUSTIACATION JUSTIFCATION

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July 26, July 26, M996 i1996 . . .

Complete ithe Root Cause/CATPR according to the action plan.

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• 11, 1996 PCAQRB Chairman . June

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NRCOOI-1642 NRC001-1642

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NRC001-1643 NRC001 -1643

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"TQ:" SYME ' ~.I. .

.Ou.:

Due: Nov.inber

.Nov mber 21; 21 : 1:997:

1997 .... . . .'" '

Action:

M~OW:' I )Pl .se ~pfete 1)PI$ase comrniplete ~e. Root-qaus~and the. Root Cause and proPCtSeanct propose and

. juitify.

justify CATPR.

CATPR.IAW*NGoNA IAW NG-NA.702. ..702. ..

2)PI.,ase obtain .Manager 2)Please Manager concurrence.

concurrence.

PCAQRB PCAQRB # qJl- $/6('0

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NRC001-1644 NRC001-1644

MtfY. REPORT (PCAGt)

L "

i im . '

• i ii i I i . .. "

. if=

8

~ ac:ceA The restricted area on the lOp access to the ansa reactor vessel head, below the CRDM top of the reactor CROM flange resulted in irisulat*on, has teaUIted ftange insulation, Inadequalt ability to c:ampIeteIJ an Ir!adequate lnbpect and clean completely tnapect dean the outslcel1op outak(eItop surfaceaurface of of the r8IICtDr v*

• head; reactor vessel head; *This surface This surface ro requires inspection/deanlng due reqUlfesinlpectionlcleanlitg potential for boric acid clue to ttie PQtentill accumulate em acid to accumulate head, as a on the head. a result System fe8Icage res",3 of :aystem leakage andthe concerns that and*the corrosion concem. ae essceiatad that are 4sscciated with acid leakage.

with boric acid ~e. .

El CONTINUED CONTINUED

0. COR!TE AcTM TO PREVENTRECURRENCE

• Modillcation Modification 94-0M5 been initiated to m&taDg 94-0025 has been Install 9 knspecftVaccess inapectlonlac:cass holes. holes, with removable covent, covers, In in the service service structure.

a~. The access* access hoIeaals will allow both dect allow bath remote Yiauallnapec::tion direct and ",mote visual Inspection capabilitles.

capabilities. The The. modification modification will also WjII also access to the top surface adequate acceu

• allow for adequate surface of the head cMan/remove any accumulated heat to dea1/remoVe accumulatedborl~ boric acid add buildup. The The

• modification has been approved for implementation implementation during 13 RFO by both the PRC and the WSC o0 CONTINUED CONTINU!ID

e. JUSTFAMIMU FORCORRECTIVE ACTON*

• 1_~IatIon.of lnsta'ion of thethe additional access holes in the service structure the.service structure wIll will provide the access that is needed needed *10: t: Inspect and and: . .

c:teM clein the top surface surfaceaof the11!llCttJr theieactor head. Inspection/cleanlng activities foIlowiJig Inspection/cleaning fllowing modlflcetlon Inlstllationwill PI'QVidEt 1JlQd1fica1iOl:J*lnel!allation:'t:ViU .provide, .: .

baseline Inormation for baSaflnelnfCrrnatlon ftre inspection for future Inspection purposes.

purposes. PrevioUs lnspetons of the Previous Inspec;tlons the :v~

vessel hed~ and an;,tlyliiaanalys of of the the*corrosion.

COf!OSIon present has determined that installatloA conditions present modification can be installation of this modilIcallon ~ scheduledfor ld)eduled1'or 13. 13.RFO RFO with very limited With very. IHniIed rtak.

risk .

of damaGe damage to the surface surface of the head from bone acid cOrrosion.

boric acid corrosion. . .

[I CONTINUED F. w* EMflON ."CU:EiMs Modification 94-.0125 CATS Item RECE.VED 7841 0 21 .. ,

F~k? -9 PCAORB.'1 1MMIT UED:

LM.Ji~oU~rrcOON~niM.

NVABE-C:NC, i .,./

J. ir* OA D*.~ - *?

-:~

I. .. . . . . . .. . . . . *.. . . .: . ." .:.,. .

1  : Y,

[ 16 6]

[166] 0 I

NRC001 -1645 NRC001-1645

~AUTykEPOftr c E~T On Thw4day, August 7 the Wow-Iidtid ptaon let to discus hs -AQ-R.

-,1974

'1 , /i Thafollowing actions were-detoinrol tobe neu to res-lvethi issu. Thesel -

actions 4m to dptermine ithe Mod will provide the ability to.dmot head whc is ementlA to an sifective evalation of any boric-acid found on thi bead. TMWls a :d

• h*+e~mc! tt+ m+f*ekm

• atotgm..

I, ppirrukpacmwcOWNflno1eke Ru"Sc h annm fimat Attendees:Prasoon Altendees: PrasoOiI (oyal Goyal John Haisrpan Jolm I!ar1igan Mike Shephord Mike'sbepliQrd Ed.Chbiahusky Ed. CbGnahusky Dan aley Dan Haley OlennMcIntyre Glenn Mcintyre wdl thethehead canbe cleaned after the installation of the 4rWger

,Determinehow 1".1'.Determine how.weli head can. be deaned after the instaUition oflbe larger other inalled the proposed Modification and winl access openings;:Three aCcessopenings.: 11uee other Units.have units have m~ t~ proposed Modification aod wiD .

enables the complete cleaning of the head.

d/.remiue ifthe contactedtoto.ctetermine bebecontacted IftheMod Modcnables the campleteDue: cleaning of~ he:&d.

November 1, 1997. .

PrasQoJiGoyal

• Action:Prasoon

.Action: Goyal . Due: 'Nov-embed, 1997.

A.-2.. M.vuaW EvaluateUm po.ibWtiesmrforaninau~liuvo thePOiMimucs aJterriailv~.bead be implemented, the-Mod

'O ng i:feaning ~"

would

. '- If. suitabJe*

£ D sOIC be iinne.,Ces"z .,

process of cleaning the bead process'ofcle8zdngthe 1ieadcan can be implem~cd, th~*ModDue: Would be ~:

Novembr -J,1997 Action: Dan -illy, Adion: DanHaleYt SYstcm*.En~

SysHem-Eygineer , . Due: NO\Itmbed.l997 Fall 97 refludaig outage.jad evalua .tehehead

.3.,3, ..Visit TATMI;ifif'J;q88ibk.

visit after the durins*~

ossible during.their installation Falf97 of larger rd4dinS accessoUtagePorts. ..ad :~_the head.

inspectiOdcapability inse.tion capability:ifter thC instaJlaiiODOfJar8cr.~ ports; .'. oebr1,." .u ' .

1997, Action:Prasoon Action: (Joya, D)ENS PJ880on~l>EMS' '..' . Due:' NOVember 1.1997 .. '

4..4..Determine ifthe Modificatioiai requi**., and ~ eat to the P*G1orifp.Fced.dure.,

~etermineif~ModifiCati~i~:I'eq~and to ~ ~G.;o~:ifpr<<ed1itc* requiired'. .' .....::

0vaions revisions are, required tocaiyepcatin orteEgioun rmaio.. -="ereq~ire~ho ~ ~OIlS ror.dle~8~~oriS.~.:: ::;.

• byNG-VN-)324, Boric Acid.COMiaioý.-. .. ',

• b}iNG-$N-00324~*BoricAdd:CorrcmoiL:  : .... ' . ,

Due: Veci.mberI , 17 .

Action: Dan Ha-,System Engineer ::

• AdiQn.:**DanHU,iy;Syuem.&gilteer ~:~l; 1997.:.. ' . : ?, ,:

.The1he.Ii~

5.1

• i.t iti-r action ac:tiOnis: eD~:ontJiubove is$iondn PresatXWoto.PRG on'the above a*tui ..cruoa,:

s.

S* .Eiu.r* PteIeirt MOd to:PIlG* .' '. , .. . . ' ";.'

• ~::*~.syStenis:En* ...* , ,'Due:"!~ wy2899 2S:'l99i- . :

. ... ';/.~'.. ' ~' .. ',: .: .... : .. : .:~ '. '.::;"L:~: .':

. Or Imptnemei-EnhancedClaning Process.

Or* 1~*EnbaDccdCJ~~* ..

Ad,i<t~: M.-dawc

',~~calS~eIIIS' ~:*~},~~..:l~.. :

Action: S.i W.b.n . " ... e": sr28,198s Or '"

Or ;n. N*-,.,EN-00324"- '.'

llevitcNG..BN..oo324* . '. ., 7]

Mechancal SystemslEngmneenag.Apri.- Due:

htion(M~cal*SY_5~~.

A.tion Due:.' ApriU;l9p~': .

I 6

Supervisor, Mechanic al, System. ........

I.

rt I.

'+yr ii..

[:L

[16 7]

[l67]

NRCOOI-1646 NRC001-1646

QU~flEPVOlTCONrTM IV1O44RUT 641 WWI

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tus[7. . ' : .. ** i*:'"

• _I*ilI:d***~r1n.~*~~t~0i~~t4~:*~:i}**'~. .

The ofo~igaxeM~ s respoasu.o he ssl.edhem on d.ei*oce itcx.- sw o

_1

• i : '; .

2L1.EBvE. . alulm

. . . .dan alternatlvv head

~ IlllIIIerDaIhe demnlg proem.

baddeaDJD. proc:as; Dte to dtiheinted accesS to th

~e to die limited aa:esa to 1M.

surface 4f the reactor vesn head. both visually and physically and todth contnaints placed surface 4fthc reactor, vessel head.both viS1l8lly andpbysically and l(J'~ C:onauai~ placed onqnthe i*pection/clcanlng activitids by schedule and ALARAnonew ble methods. of+.s lhe iqspectionldCllllhlgactivities by schedule and ALARAilo~feui~re ~ d(:

inspectionand inspection andororce aning have been folltld. DisculIIiilina wit;b other PWRsites

~hig have Eetn found. DiscussioIs with other PWR .a revealed skea fJaJ revealed' .

that most have already modified/enlarged theaccss openings through the strvice structun to that.most hw IlIready inoditiecllen1arscd tbcICCCS$openinp through the.serviee St~ 10 the upper head surface or are planning a modif'cation to enla the the upper head SUrf8llC or are planning a modiflCation io enlargetbe ~lopenin.JS; ~n the . access openinsS, n the serviccistructure.

service structure*

4....Determinm dIOClJicatioaIsIsrequired DetermineNII'a *.modimcatlon procedurestotodclary reviseprocdures reqidredorOfrevise eIaiIfycurrentcarnnt

[processes. Without access to the entive suriace of

.p~. Without ~ess to theentiresutface of the ~tor Y.l~*it is ~possible.the reactor vessel head/t is " possible to10perform inspectionand completeinspection perfonnaItcomplete the bead.

cleartingofoflile andcleaiting Th.. limited*insp.ction.

~.::'I1lc*;~~~on:. ".

me existing' bodo.cystal depositson the.

capabilities thatexist capabilitiesthat shownthat haveshown exiithave there.2mr thatthem arescn:ne exjstiDibOroa' ~~ *~ts.cm the*

swfac:e.*These headsudrace.

head access holes. W,

.bed:

dposits cannot hebereachedb foir

~c.dcpOsitsQlJlnot fOrc~lea ninjWr'wovel hroghdiecuren c)~~o,,~~.ibJ~f:um;~*

~ et"Whs I

=~::':=:::f=~~~!~=:=tr=

Modifying Modifyingthe ~.~

lbespot" holeswould ac:c:csSholes structure access would allow aJiowfOr COinpJ~teinspeclion (dra.*compiete of "he ..

inspi!Cti~:of:"the:

clea the surface of the heia to remove any exisft borw ryta...

head surface, aces to heid surface. acccssto c:1_lhe surfai:c of thein'eco ~ t'" temO~ . ag.i;,ue bor:onaystai*.*

aIJY'cxiatins sYME.

Sties, depo$ilsand depoýits shouId.allowfor andshould.allow improvedfutqm forimproved fu~ in~~lig.~~e.ti1Des..::SYME. ..' .,.1:

• dftermincddtht hoa dbtumined sJiOuJdbepursued.

should thatOhU1emodification be pursUcci...

modification.o to enlarg theacces ho

~nJ.8rgetheaccess*hC,tleS:in hd support

.t the*~*

. . . . . . . . ' .. : .":::.:", '. '.~:' ":. ~~::

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NRC001-1647 NRC001-1647

.. .: ': 964 51'- . _ .

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.. .:',SYME~

To Y .': ., " ~.: ,.: :' , .. ,'" .' ,~ .. ,,: ..... :,

"D~e:  : Decembert1.:19gfJ,

" DUe:DeoeMber 11, 1998 ,:  ; "l '

• ,Action:~ Th0Part.6 res~e

• Actlon:* i.1;1l1EtPart.$ s** e wMs rejected by the: SRB/PCA.QRB was~telected t)y.theSFlBlPCAQRB: . '

bec8uae ftIt didn't because ~dn*t include InClude the required attuibutes the required atb'Ibutes. or of. a full'fu1IRQot Root ,', :"

Cadse EvaluatiOn (e.g.,

Ca&Jse Evaluation problem statemnent, (e.g.,problem statement" event eventn..-ative, norrative, why. why;;*; .,',: ",

prgfious correctiveactionsl.I' assessm~ wereJn8ff~ve: to prerfous corrective actions/self assessments wereIneffectve t(): ': ~ ,

rec~ence *. etc. --see:

pre;ent recurrence, pr.eent NG-N.A-702 step see NG-,NA-702 ~ 6.7.2.b). 6;~.2~b)~ , ' , :.',

complete aa Root

2. Please complete C,ause Evaluation'lAW Root Cause Evaluation 'IAW NG-NA-702. NG~*702.

3. Please Plese address address the 'Oavis-Besse response the Davis-Besse response to G(L.88-05 GL:8&.-05 for the evatuatlon of evatuation the. sig~lflcalJC8of Of the significance of boric boric acid on the tl:t~ head. ~ead.' ,

obtain"manag.r*concurrence~

4. Please obtain.manager concurrence. ... ...

SRB/PCAQRB October 28,-1998

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NRCOOI -1648 NRC001-1648

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ADVD8I TO ADVER TO qUAMiT WONM,(PCAOMI qUALITY 1IIPOR1'(PCA4R) 19608q,

[3 CaeMEAt Other ProceIa oc NOW-nSME 40 Wat1l1 Action Docwnent RECEIVED o CategOty 4 Document InI_ted Chaligs to MRC ChMg& MAC Recommendallbn Recomimendmfibn W11 Catepy3

~ Categoly2 CATPR Catogoiy3 CATPR ONO ONO 0rgJYESYES PAR Category 1 C, UE DATI pan Part 5 1' I I Part 6610115198 1iO~l5190 I'. m.

I Thi* PCAQR CIt.

-*f PCAQR Part 410 Cau... The condition Identified a Root Cause.

potentially adequately view portions potentially not adequately propose an Apparent submitted to propose 4 Is being aubmlttedto kMntJfled by this PCAQR

¥Heel head reactor vessel portJona of the reactor Cause Root Apparent Cau. Root Cause PCAQR was that an engineer engineer could could Instead Cau. . lnatelld evaluate for Boric Acid

...... to evaluate AcId Corrosion c:oncem.~

I Convalon concerns. The depth d8pth of Investigation determine.why he could not ....

Investigation to determlne.why see doe8 does not not warrant Ii wanant ROOT CAUSE.

a ROOT couldn't view an CAUSE. He couldn' all of the vessel oftha YeaHI head .... becauae the access area because 8CC888 las to do .0 baIIn so . are . not enough to aI'ow nat big enoUGh it. The attributes allow It. completion of *a attributes to be gained by completicn full blown Root ea full Cause.... wltl will not Ghang_ corrective actions, change the comlCtlve ac:tIona, which am to make bigger which .... hole.

bigger hotee or more adequatllly 01' adequately utiltae exladng holea.

uti... the existing holes. An eventevent nardve.

naratdve, extent condition, and extent of conditIon, and

pnwIouaocc:urrence previous occurrence eYllluation evaluation will not add any obMrvable Ita.. situation. This observable quality to this this

....... 8dd....... the softWare Ussue addresses software IIssue

..... of Inspection, lnapec:tlon. not the hardware ..... of head leakage.

h.rclwmelIssue

An apparent cau An appllrent cause . . wlllntO" will more tMn than adequately 8Upport Corrective Action to Prewnt support ComtCtive Prevent Occurrence.

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I o3 CONTINUED DCONTINUeo WA o0 CONTINUED CONTINUIED 94O EVALUAOR U~l SURVELANCE PHCA NO DATE a ~yj~ .8310 01-19-99)

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NRCOOI-1650 NRC001-1650

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z ONSION EXTECAEGRY REQUJESTED UNTfL (SeeNGN4~Aftdhme 4)

-Y/2 7 7

DREASON FOR EXTENSIONEQUEST .

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0 CONTINUED E. USTMCATION FOR DELAY (ADDRESS EFFECT ON HARVARE,.PROGAS, AND ACTIVITIES)

~on~.--of Or4L.C,,'tn '.AA i'~er'n.

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[172]

NRCOOI-1651 NRC001-1651

.'" ".: i 0o CONTINUED CONTINUED Due IJ12U91 UNTIL D.REASON FOR FOR EXTENSION REQUEST suf&iCOltim*t Providesufficut Provke time to attached action complete the attached o conpcticthe planwhich action plan whH:hwill W1111cad ddermiaation.ofwhether lead totoaadetermiuation.of whether orornotnot aa modificamion IIIQdification be pursued shouldbe should pumItJd or or tMe theprocedure rmsect or proccdllre Telscd potI:IIlWly an orpotentially cptionto an option tDperform perform an cleaning without enhanced cleaning ancnhNced withoulihcthe Mod.

Mod.

0 CONTINUED 0QCONTINUED E. JUSTIFICATION FOR DELAY (ADDRESS EFFECT ON HARDWARE, PROGRAMS. AND ACTIVmEIS)

Then=

The iospcclionwill DCXl inpection IIO'l be will not pcrtbnncduntil be pesfdbru .ITF.

1IIItil.1ZRFO. This Thisdeay win provide delay will provl.Je additional IIIIIIce aa more inlOrmatio~ to make additional infdrmafioito mcm: inormeal infbrmed decision without

• . decision implementation. Based impIctiugimplementation.

witbout impacting BasccI on OIlthc tIJe input fIomthe involed input fwam.the involved engineers engiDCtl'lIhis a:ccptablcand this isisaeptable and isis aa delay or delay of

.oly lmoath. There

,aaIy Imcmii,. MIl be There vill be no impact on 110 impact Oil hardwae programs or banlwarc. programs activ'ties.

oractiviticis.

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[181]

NRC001-1652 NRC001-1652

e FIGURE 1-1 FIGURE I-I B&W-DESIGN CRUM NOZZLE BlW*DESIGN CROM NOZZLE

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NRCOOI-1655 NRC001~655

SEV'ERJl\L D., Trl="'* Of BORIC AOO

't/'ERE fOUND ON THE REACTCR NRCOOI-1656 NRC001-1656

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ElCONTINUED ELOseANImuSO DATE CATEARCA 50fY 2-March 20. 199 Man:h 20.1&198 The Z197 extension request action plan has raac:h8d reached-a dacislon Point 11 decision point to proceed proceed with wtth a propoeaIlo modiy the rebactr proposalto modifytl'ie reactQr '

....!..,nI_ structure.

headsupport 1t.. structum. ThIs.xtension This.extension wIRwil agow time for Me aIlowtln1e process to ~

modifcation pmcesa the modification, ~and be stated and aa:.~aIIOn pswtalbon to the the IPRIG. schedulEd. PRG PRG scheduled. PAG appl'OY8l approval of ,the the modiflc:atlon,wHl,iubstantlate the moditlcation Sa modificatlon .will.ubstantiate the'modlflcatlon CAT~FUotihis as aa CA'PR for this -P.CAOR.

PCAQR.. ", '.

0CONTINUE0

.AiSfWCAMFlONFO DELAY (ACOASSS EffETQWNAFWWAFM. PROOOGAM. ANUA~nM1IB5 l a non reportabie condition and not a ~non confovnc with meo'k t to te atr headl..te.bofiriccd~ooirroio Is pmcus Isa tobw acting paices and does rot pose any Imamedia safetyconcern for the rwfitrshor tr'm.

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~roatc:.u..;CATPR~ObIaIn:M~coricurreriC8 loplet root cause. CATPR. arop obain. Manage conicurence {lA>.Jl 4a-% ~.

4MC&04- ~ 3~J/)-q~

k*LA  ;..j 0 q( -

. .1 01-ORGANUWnON OWDAW am TEMOR MgMERM UNTL 15,1998 October R

R A A decisiOn decision POlr1thas POWn has been reached to toproceed proceed with a8 modlfr the reactor.head proposal to modft.

a propcisel reactor head 181Yiceatruc:ture.

seivice hrutnUre A A e ~oflhe u

PmTu ftn~al proposed ~

of the PfOPQUd modification was lildUded Included _on On N the agenda t ' " forfr the th PRGrneetIng:lbiit PRG maeeing ifat warn scheduled for .-tIChedutect meefing on 3I03/9Jl-~- ~ Gou/dnotbe_

could not be ~~ pmesen'ted. ;_~~RFOj1 Q

.U 3n3I9a; The PRG

&OWN19. The PRG meetlng.on 31019 was conpulled and and thefthproposal D~tORFO I1I -

-E- ~adule-conatralnts.

-f ichedle constraints the PRG wit the wit: not meet aplnd not viles agadn unt 8fter aft

~. ~e.--

the outaget The

,2098Duef The'-3120i98:due ~-;b,:-.IS:P.CARC:l*c8n daft" tohbSPAR a -not o' --

8 presentaon after the reuln.ur~

T T *WNmet.

.. h extension request met,the~alon reques will tIine to allow wmn will allow schdule the to schedule the presentatiol1~the

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Ir lrfý W UIM MCA UN UAY ymna FIJ wha On ?UU3 P!

-TNa:**aenon

-ThIP non re~condItJonand reportable ,condtn and not a~'~lVitlr~-tO~rNctOrhe8d, bafornc with MeMad to 0 reacr hea8. -n.~:~d~- Tha bo0:id o msA i . ' . - ' . . . '. . ". " .. " .,' .. ,' ........ ', ,'....,'

rocse

-POC8III is a sowacting proess isa lIkiw acting proceM ai1d doe$.i1qt snd does not PQSO:any po -,any immedite

~Iate safety ~ tor.

$lrely conc o ~ ....

the nea,. shotter.m hof:t~-

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